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Sample records for bwr stability operating

  1. BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-07-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  2. Stability prediction of continuous surveillance in BWR reactor; Predictor de estabilidad para la vigilancia continua de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tofino Gomez, Y.

    2006-07-01

    As result of the susceptibility of the Boiling Water Reactors (BWR) to suffer from power instabilities, the program LIP has been developed (LAPUR Input Preprocessor), which automatically determines the decay ratio (DR), as stability margin indication. For DR calculation, LAPUR program is a good predictive alternative: a fast execution for an acceptable precision. LAPUR demands a complex input, dependent on the instantaneous core configuration, requiring an exhaustive control of its generation. LIP, with a modular character, automatically generates the input from the core monitoring system, CAPRICORE (based on Simulate-3), obtaining the DR during the operation. This tool can accelerate the start-up maneuvers and other transients, increasing the plant availability. (Author)

  3. Assessment of the Prony's method for BWR stability analysis

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    Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico); Castillo-Duran, Rogelio, E-mail: rogelio.castillo@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico); Palacios-Hernandez, Javier C., E-mail: javier.palacios@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico)

    2011-05-15

    Highlights: This paper describes a method to determine the degree of stability of a BWR. Performance comparison between Prony's and common AR techniques is presented. Benchmark data and actual BWR transient data are used for comparison. DR and f results are presented and discussed. The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

  4. Predictions by the proper orthogonal decomposition reduced order methodology regarding non-linear BWR stability

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    Prill, Dennis; Class, Andreas G. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). AREVA Nuclear Professional School (ANPS)

    2013-07-01

    Unexpected non-linear boiling water reactor (BWR) instability events in various plants, e.g. LaSalle II in 1988 and Oskarshamn II in 1990 amongst others, emphasize the major safety relevance and the existence of parameter regions with unstable behavior. A detailed description of the complete dynamical non-linear behavior is of paramount importance for BWR operation. An extension of state-of-the-art methodology towards a more general stability description, also applicable in the non-linear region, could lead to a deeper understanding of non-linear BWR stability phenomena. With the intention of a full non-linear stability analysis of the two-phase BWR system, the present paper aims at a general non-linear methodology capable to achieve reliable and numerical stable reduced order models (ROMs), representing the dynamical behavior of an original system based on a small number of transients. Model-specific options and aspects of the proposed methodology are focused on and illustrated by means of a strongly non-linear dynamical system showing complex oscillating behavior. Prediction capability of the proposed methodology is also addressed. (orig.)

  5. Full scale stability and void fraction measurements for the ATRIUM trademark 10XM BWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Wehle, Franz; Velten, Roger; Kronenberg, Juris; Beisiegel, Achim [AREVA NP GmbH, Erlangen (Germany); Pruitt, D.W.; Greene, K.R. [AREVA NP Inc., Lynchburg, VA (United States); Farawila, Y.M. [Farawila et al., Inc., Richland, WA (United States)

    2011-07-01

    This paper describes recent advances in BWR fuel testing at AREVA NP's KATHY loop including stability and void fraction measurements. The stability tests for the ATRIUM trademark 10XM bundle with corner PLFR's were expanded in scope compared with previous campaigns to include simulated reactivity and power feedback essentially reproducing BWR operational environment. The oscillation magnitude was allowed to grow to explore inlet flow reversal and cyclical dryout and rewetting. The void fraction measurements employed a gamma ray computed tomography technique that reveals not only the average but the detailed sub-channel void distribution, and the range of measured void fraction has been expanded to higher values than was previously attained. With the completion of the required licensing tests and stability performance demonstration, the ATRIUM trademark 10XM is available and fully qualified for reload supply. (orig.)

  6. Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

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    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

    2006-07-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

  7. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  8. Prediction of the stability of BWR reactors during the start-up process; Prediccion de la estabilidad de reactores BWR durante el proceso de arranque

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R. [ININ, Km. 36.5 Carretera Mexico-Toluca, 52045 Salazar, Estado de Mexico (Mexico); Blazquez M, J.B. [Centro de Investigaciones Energetics, Medioambientales y Tecnologicas, Av Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  9. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  10. Development of Sirius facility that simulates void-reactivity feedback, and regional and core-wide stability estimation of natural circulation BWR

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, M.; Inada, F.; Yasuo, A. [Tokyo Electric Power Co., Inc., Central Research Institute (Japan)

    2001-07-01

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of fluid conditions, power distributions, and fuel rod thermal conductivity time constants, including the normal operating conditions of a typical natural circulation BWR. The results showed that there is a sufficiently wide stability margin under normal operating conditions, even when void-reactivity feedback is taken into account. (author)

  11. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

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    Ilas, Germina [ORNL; Betzler, Benjamin R [ORNL; Ade, Brian J [ORNL

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  12. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

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    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  13. Performance of iron-chromium-aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    Science.gov (United States)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun; Heuser, Brent J.; Mandapaka, Kiran K.; Was, Gary S.

    2016-03-01

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

  14. Development and Testing of CTF to Support Modeling of BWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gosdin, Chris [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

    2016-01-29

    This milestone supports developing and assessing COBRA-TF (CTF) for the modeling of boiling water reactors (BWRs). This is achieved in three stages. First, a new preprocessor utility that is capable of handling BWR-specic design elements (e.g., channel boxes and large water rods) is developed. A previous milestone (L3:PHI.CTF.P12.01) led to the development of this preprocessor capability for single assembly models. This current milestone expands this utility so that it is applicable to multi-assembly BWR models that can be modeled in either serial or parallel. The second stage involves making necessary modications to CTF so that it can execute these new models. Specically, this means implementing an outer-iteration loop, specic to BWR models, that equalizes the pressure loss over all assemblies in the core (which are not connected due to the channel boxes) by adjusting inlet mass ow rate. A third stage involves assessing the standard convergence metrics that are used by CTF to determine when a simulation is steady-state. The nal stage has resulted in the implementation of new metrics in the code that give a better indication of how steady the solution is at convergence. This report summarizes these eorts and provides a demonstration of CTF's BWR-modeling capabilities. CASL-U-2016-1030-000

  15. Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  16. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  17. Detecting long-range correlation with detrended fluctuation analysis: Application to BWR stability

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Apartado Postal 55-534, Mexico, DF 09340 (Mexico)]. E-mail: gepe@xanum.uam.mx; Alvarez-Ramirez, Jose [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Apartado Postal 55-534, Mexico, DF 09340 (Mexico); Vazquez, Alejandro [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Apartado Postal 55-534, Mexico, DF 09340 (Mexico)

    2006-11-15

    The aim of this paper is to explore the application of detrended fluctuation analysis (DFA) to study boiling water reactor stability. DFA is a scaling method commonly used for detecting long-range correlations in non-stationary time series. This method is based on the random walk theory and was applied to neutronic power signal of Forsmark stability benchmark. Our results shows that the scaling properties breakdown during unstable oscillations.

  18. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

    2003-07-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  19. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H. P

    2002-03-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

  20. Aging and defect characterization of motor-operated valves: progress based on NPAR strategy. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Eissenberg, D.M.

    1984-01-01

    The Nuclear Plant Aging Research (NPAR) program strategy is directed at carrying out comprehensive aging assessments in order to define and resolve issues related to aging (including service wear) of electrical and mechanical components and structures at operating reactor facilities and their possible impact on plant safety. This paper describes work recently completed at Oak Ridge National Laboratory which applied the NPAR strategy to motor-operated valves (MOVs). The objective of the work was primarily to develop an understanding of the operating history and conditions and the failure modes of MOVs in nuclear plant service as a preliminary to identifying and recommending methods for trending aging degradation. A second objective was to demonstrate, using MOVs as an example, that the NPAR strategy can be applied to many electrical and mechanical components of nuclear power plants.

  1. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  2. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  3. A novel auto-correlation function method and FORTRAN codes for the determination of the decay ratio in BWR stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Behringer, K

    2001-08-01

    A novel auto-correlation function (ACF) method has been investigated for determining the oscillation frequency and the decay ratio in BWR stability analyses. The report describes not only the method but also documents comprehensively the used and developed FORTRAN codes. The neutron signals are band-pass filtered to separate the oscillation peak in the power spectral density (PSD) from background. Two linear second-order oscillation models are considered. The ACF of each model, corrected for signal filtering and with the inclusion of a background term under the peak in the PSD, is then least-squares fitted to the ACF estimated on the previously filtered neutron signals, in order to determine the oscillation frequency and the decay ratio. The procedures of filtering and ACF estimation use fast Fourier transform techniques with signal segmentation. Gliding 'short-time' ACF estimates along a signal record allow the evaluation of uncertainties. Some numerical results are given which have been obtained from neutron signal data offered by the recent Forsmark I and Forsmark II NEA benchmark project. They are compared with those from other benchmark participants using different other analysis methods. (author)

  4. Effect of a Chloride Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 2)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2002-11-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the second test of working package (WP) 3 with a NaCl transient, performed at Paul Scherrer Institut (PSI). In the first part of the experiment, an actively growing EAC crack with a crack growth rate (CGR) in the range of the 'low-sulphur SCC line' of the GE-model was generated by periodical partial unloading (PPU) in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm). Then a chloride transient of 49 ppb Cl{sup -} was applied for {approx}40 h. After this transient, the load level was reduced and the loading conditions were changed to pure cyclic loading. Thereupon a second transient with a chloride concentration of 49 ppb was applied. In both RPV steels, the first chloride transient of 49 ppb Cl{sup -} resulted in an acceleration of the EAC crack growth by more than one order of magnitude and in fast, stationary SCC crack growth during the constant load phase of the PPU cycles at K{sub I} values < 60 MPa.m{sup 1/2}. 3 h after adding chloride to the high-purity water, the EAC CGR started to increase in the high-sulphur RPV steel during the constant load phase of a PPU cycle and after 20 h a stationary EAC CGR value in the range of the 'high-sulphur SCC curve' of the GE-model was reached. After 5 h in high-purity water, the crack growth began to slow down after a partial unloading cycle and 15 h later it reached again a stationary CGR value in the range of the 'low-sulphur SCC curve' of the GE-model. The second chloride transient did not result in an acceleration of the crack growth in both investigated specimens. This was explained by crack closure effects

  5. Results of VGB research work with respect to operation of BWR pipes made of austenitic SS; Ergebnisse des VGB-Forschungsvorhabens zur Absicherung des Betriebsverhaltens austenitischer Staehle in SWR-Rohrleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Kilian, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Bruemmer, G. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany)

    1998-11-01

    The VGB research project was to examine and characterize various, operation-induced impacts on the crack formation in stabilized austenitic steels, caused by intercrystalline stress corrosion cracking as a result of sensitization after chromium depletion at the grain boundaries. The results of this project as well as available operating experience show that the measures taken so far for the future operation of the German BWR plants, for avoiding in these plants intercrystalline stress corrosion cracking, correspond to the state of the art and achieve the wanted purpose. These measures are: use of optimized material W-No. 1.4550 with reduced carbon contents; use of optimized welding techniques for reducing the heat input and the welding shrinkage (cold deformation.); optimized preparation of welding work in order to avoid shape defects during welding (eg. edge misalignment, defective mash welds); reduction of tensile stresses occurring during welding; compliance with the recent VGB water chemistry code. (orig./CB) [Deutsch] Das VGB-Forschungsvorhaben sollte verschieden gelagerte Einfluesse auf die Rissbildung im Betrieb von stabilisierten austenitischen Staehlen, verursacht durch interkristalline Spannungsrisskorrosion infolge Sensbilisierung durch Chromverarmung an den Korngrenzen, systematisch erfassen. Aus den Forschungsergebnissen dieses VGB-Programms sowie den bisher vorliegenden Betriebserfahrungen ist festzuhalten, dass die bisher durchgefuehrten Massnahmen fuer den zukuenftigen Betrieb der deutschen SWR-Anlagen zur Vermeidung von interkristalliner Spannungsrisskorrosion zielgerichtet waren und dem heutigen Wissensstand entsprechen. Diese Massnahmen sind: 1. Einsatz von optimiertem Werkstoff W.-Nr. 1.4550 mit abgesenktem Kohlenstoffgehalt; 2. Einsatz von optimierten Schweissverfahren zur Verminderung der Waermeeinbringung und zur Verringerung des Schweissschrumpfes (Kaltverformung.); 3. Durchfuehrung einer optimierten Schweissnahtvorbereitung zur Vermeidung

  6. Model for the analysis of transitories and stability of a BWR reactor with fuel of thorium; Modelo para el analisis de transitorios y de estabilidad de un reactor BWR con combustible de torio

    Energy Technology Data Exchange (ETDEWEB)

    Nunez C, A. [CNSNS, 03020 Mexico D.F. (Mexico)]. E-mail: anunezc@cnsns.gob.mx; Espinosa P, G. [UAM-I, 09340 Mexico D.F. (Mexico); Francois L, J.L. [Fac. de Ingenieria, UNAM 62550 Jiutepec, Morelos (Mexico)

    2004-07-01

    In this work it is described the thermo hydraulic and neutronic pattern used to simulate the behavior of a nucleus of thorium-uranium under different conditions of operation. The analysed nucleus was designed with base to assemblies that operate under the cover-seed concept. The pattern was proven to conditions of stationary state and transitory state. Here it is only presented the simulation of the one SCRAM manual and it is compared in the behavior of a nucleus with UO{sub 2}. Additionally one carries out an analysis of stability taking into account the four corners that define the area of stability of the map flow-power and to conditions of 100% of flow and 100% of power. The module of stability is based on the pattern of Lahey and Podowsky to estimate the drops of pressure during a perturbation. It is concludes that the behavior of this nucleus is not very different to the one shown by the nuclei loaded with the fuel of UO{sub 2}. (Author)

  7. The doctrinal basis for medical stability operations.

    Science.gov (United States)

    Baker, Jay B

    2010-01-01

    This article describes possible roles for the military in the health sector during stability operations, which exist primarily when security conditions do not permit the free movement of civilian actors. This article reviews the new U.S. Army Field Manuals (FMs) 3-24, Counterinsurgency and FM 3-07, Stability Operations, in the context of the health sector. Essential tasks in medical stability operations are identified for various logical lines of operation including information operations, civil security, civil control, support to governance, support to economic development, and restoration of essential services. Restoring essential services is addressed in detail including coordination, assessment, actions, and metrics in the health sector. Coordination by the military with other actors in the health sector including host nation medical officials, other United States governmental agencies, international governmental organizations (IGOs), and nongovernment organizations (NGOs) is key to success in medical stability operations.

  8. Stability of (p, Y)-Operator Frames

    Institute of Scientific and Technical Information of China (English)

    Zhi Hua GUO; Huai Xin CAO; Jun Cheng YIN

    2011-01-01

    In this paper we study the stability of (p, Y)-operator frames. We firstly discuss the relations between p-Bessel sequences (or p-frames) and (p, Y)-operator Bessel sequences (or (p, Y)-operator frames). Through defining a new union, we prove that adding some elements to a given (p, Y)-operator frame, the resulted sequence will be still a (p, Y)-operator frame. We obtain a necessary and sufficient condition for a sequence of compound operators to be a (p, Y)operator frame. Lastly, we show that (p, Y)-operator frames for X are stable under some small perturbations.

  9. The 800-Pound Gorilla and Stability Operations

    Science.gov (United States)

    2010-06-30

    THE 800-POUND GORILLA AND STABILITY OPERATIONS by James P. Hunt, Major General, USAF Deputy Commanding General, MNC-I, April 09 – Jan...control number. 1. REPORT DATE 2010 2. REPORT TYPE 3. DATES COVERED 00-00-2010 to 00-00-2010 4. TITLE AND SUBTITLE The 800-Pound Gorilla and...national interests in this ever-changing world; each must change and adapt to succeed in Stability Operations. …WHERE DOES AN 800-POUND GORILLA SIT

  10. Operationalizing Economics for Counterinsurgency and Stability Operations

    Science.gov (United States)

    2009-05-01

    easily adapted to those types of operations. Development is a complex problem and is central to counterinsurgency and stability operations. A crisis ...Joint Operations, Vol. 1. Washington, DC: Government Printing Office, 2006. Heybey, Berta and Lisa Bush. Do Crisis Response Operations Affect Political...3 (March 1998): 16-9. Sen, Amartya. “Behavior and the Concept of Preference.” Economica 40, no. 159 (August 1973): 241-59. Shircliffe Jr., James

  11. Procedures for using expert judgment to estimate human-error probabilities in nuclear power plant operations. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Seaver, D.A.; Stillwell, W.G.

    1983-03-01

    This report describes and evaluates several procedures for using expert judgment to estimate human-error probabilities (HEPs) in nuclear power plant operations. These HEPs are currently needed for several purposes, particularly for probabilistic risk assessments. Data do not exist for estimating these HEPs, so expert judgment can provide these estimates in a timely manner. Five judgmental procedures are described here: paired comparisons, ranking and rating, direct numerical estimation, indirect numerical estimation and multiattribute utility measurement. These procedures are evaluated in terms of several criteria: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Situational constraints such as the number of experts available, the number of HEPs to be estimated, the time available, the location of the experts, and the resources available are discussed in regard to their implications for selecting a procedure for use.

  12. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  13. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: augusto@correo.unam.mx

    2004-07-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  14. OPERATING STABILITY OF MINERAL WOOL PRODUCTS

    Directory of Open Access Journals (Sweden)

    Perfilov Vladimir Aleksandrovich

    2016-03-01

    Full Text Available Creating an effective insulation envelope of the building is possible only using high-quality materials, preserving their characteristics both in the early stages of operation, and for the whole billing period. It is an important opportunity to assess the thermal insulation properties and predict its changes over time directly in the conditions of the construction site. The products based on mineral fibers (rock and glass wool, basalt fiber are the most widely used type of insulating materials in the domestic construction. Therefore, the operational stability valuation methods must be primarily created for this group of products. The methodology for assessing the thermal insulation properties includes two main components: testing equipment and methodology for assessing the operational stability. The authors tested the methodology of the accelerated testing and prediction of durability for mineral wool products of laminated, corrugated and volume-oriented structures. The test results give good convergence with the methods recommended by the building regulations. Application of thermal insulation materials are an effective way to form the thermal envelope of the building, reducing energy costs and increasing the durability of building structures. The material properties are determined by their structure, which is formed during the technological impacts.

  15. Effect of Loading Transients on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Tests 3 and 4)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2003-04-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the third and fourth test of working package (WP) 3 with loading transients, performed at Paul Scherrer Institut (PSI). Two different low-alloy steels (20 MnMoNi 5 5, 0.015 wt.% S and 22 NiMoCr 3 7, 0.007 wt. %S) were investigated in oxygenated high-temperature, high-purity water (T = 240 {sup o}C, DO = 400 ppb) in a daisy chain at two different load ratios (R = 0.8 and 0.2). In the first part of the experiments, asymmetrical saw tooth loading with different rise times {delta}t{sub R} of the load and different loading frequencies were applied. Then the loading conditions were changed to an asymmetrical trapezoid waveform loading (periodical partial unloading, PPU) and the hold time {delta}t{sub H} at maximum load was varied. In the final phase of WP 3 PSI tests 3 and 4 the SCC behaviour was investigated under constant load. With decreasing loading frequency the corrosion fatigue (CF) crack advance per cycle {delta}a/{delta}N{sub EAC} of material A increased. Sustained EAC crack growth could be maintained down to low frequencies of 10{sup -5} Hz. The time-based crack growth rate (CGR) da/dt{sub EAC} decreased with decreasing frequency. In material B no effect of the loading frequency could be resolved. Up to a hold time of 1 h at maximum constant load the CGR da/dt{sub EAC} seemed to be independent of the hold time. Above hold times of 1 h the CGR decreased and dropped down to CGR values in the range or below the BWR VIP 60 SCC disposition lines. This behaviour was observed in both investigated materials. The cycle-based CGR {delta}a/{delta}N{sub EAC} remained approximately constant with increasing hold time. The

  16. Remarks on boiling water reactor stability analysis. Pt. 2. Stability monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Lange, Carsten; Hennig, Dieter; Hurtado, Antonio [Technische Univ. Dresden (Germany). Chair of Hydrogen and Nuclear Energy; Schuster, Roland [Kernkraftwerk Brunsbuettel GmbH und Co. oHG, Brunsbuettel (Germany); Lukas, Bernard [EnBW Kernkraft GmbH, Philippsburg (Germany). Kernkraftwerk Philippsburg; Aguirre, Carlos [Kernkraftwerk Leibstadt AG, Aargau (Switzerland)

    2012-12-15

    In part 1 of this article we explained the partly relative complex solution manifold of the differential equations describing the stability behaviour of a BWR, in particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points are of interest from the operational safety point of view. The part 2 is devoted to the surveillance of the stability behaviour. We summarize some stability monitoring methods and suggest to support stability tests by RAM-ROM analyses in order to reveal in advance the stability 'landscape' of the BWR in a parameter region high sensitive for appearing of linear unstable states. The analysis of an especial stability test, performed at NPP Leibstadt (KKL), makes it clear that the measurement results can only be interpreted by application of bifurcation analysis. (orig.)

  17. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  18. Remarks on the stability operator for MOTS

    CERN Document Server

    Senovilla, José M M

    2012-01-01

    Small deformations of marginally outer trapped surfaces (MOTS) are studied by using the stability operator introduced by Andersson-Mars-Simon. Novel formulae for the principal eigenvalue are presented. A characterization of the many marginally outer trapped tubes (MOTT) passing through a given MOTS is given, and the possibility of selecting a privileged MOTT is discussed. This is related to the concept of `core' of a black hole: a minimal region that one should remove from the spacetime in order to get rid of all possible closed trapped surfaces. In spherical symmetry one can prove that the spherical MOTT is the boundary of a core. I argue how similar results may hold in general spacetimes.

  19. Design of a redundant meteorological station for a BWR reactor; Diseno de una estacion meteorologica redundante para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, R.; Celis del Angel, L.; Bucio, F.; Rivero, T.; Palacios, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: ramses@nuclear.inin.mx

    2008-07-01

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  20. The Development of Nufreq-N AN Analytical Model for the Stability Analysis of Nuclear Coupled Density-Wave Oscillations in Boiling Water Nuclear Reactors.

    Science.gov (United States)

    Park, Goon Cherl

    A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). The model accounts for phasic slip, distributed spacers, subcooled boiling, space/time -dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation. In its final form, this model constitutes a multi -input, multi-output(MIMO) linear system, which features a general nodal neutron kinetics model. Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions. The stability characteristics of a typical BWR/4 has been investigated with the Nyquist criterion. The computer implementation of this model, NUFREQ -N, was used for the parametric study of a typical BWR/4 and comparisons were made with existing in-core and out -of-core data. Also, NUFREQ-N was used to analyze the expected stability characteristics of a typical BWR/4. The parametric results revealed important factors influencing BWR stability margin. It was found that NUFREQ -N generally agreed well with out-of-core data. This was especially true for the predicted power-to-flow transfer function, which is the most important transfer function in thermal-hydraulic stability analysis. In the stability analysis of a typical BWR/4 it was found that it is very important to use accurate models of thermal-hydraulic and neutron kinetic phenomena. Moreover, the accuracy of the nuclear input data is extremely important.

  1. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  2. An A BWR demonstration simulator for training and developing technical staff

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  3. Stabilization and Reconstruction Operations Doctrine and Theory

    Science.gov (United States)

    2014-05-22

    Bosnia—Herzegovina, Kosovo , East Timor, and of course Iraq and Afghanistan. A notable exception is Guiding Principles for Stabilization and...Galula advises commanders not to waste resources protecting terrain that would be considered critical in a conventional fight but rather to deploy

  4. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    H. Wang

    1997-01-23

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  5. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  6. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  7. STABILITY OF WHEELED VEHICLES AS COMPLEX OPERATIONAL PROPERTIES

    Directory of Open Access Journals (Sweden)

    N. Artemov

    2011-01-01

    Full Text Available Different views on the definition of «stability of wheeled vehicles» are considered and the author’s own definition is offered. A version of the structure of stability properties as a complex op-erational property is offered.

  8. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  9. CCF analysis of BWR reactor shutdown systems based on the operating experience at the TVO I/II in 1981-1993

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy, Espoo (Finland)

    1996-04-01

    The work constitutes a part of the project conducted within the research program of the Swedish Nuclear Power Inspectorate SKI, aimed to develop the methods and data base for the Common Cause Failure (CCF) analysis of highly redundant reactor scram systems. The data analysis for the TVO I/II plant is focused on the hydraulic scram system, and control rods and drives. It covers operating experiences from 1981 through 1993. (9 refs., 9 figs., 7 tabs.).

  10. Absolute stability analysis of linear systems with Duhem hysteresis operator

    NARCIS (Netherlands)

    Ouyang, Ruiyue; Jayawardhana, Bayu

    2014-01-01

    In this paper, we investigate the stability of positive and negative feedback interconnections of a linear system and a Duhem hysteresis operator. We provide sufficient conditions on the linear plant and on the Duhem operator which are based on the counterclockwise (CCW) or clockwise (CW) input–outp

  11. Prony's method application for BWR instabilities characterization

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

    2015-04-01

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred.

  12. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  13. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  14. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  15. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y. [Hitachi-GE Nuclear Energy, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-0073 (Japan); Makigami, T. [Tokyo Electric Power Company Inc., 1-1-3, Uchisaiwai-cho, Chiyoda-ku, Tokyo, 100-0011 (Japan)

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  16. What Every Airman Needs to Know about Medical Stability Operations

    Science.gov (United States)

    2012-01-01

    organizations (NGO). Thus, military health services contribute to social , political, and economic stability. JP 3-29, Foreign Humanitarian Assistance, 17...deployed for the health protection of US forces. Developing graceful, noncontroversial rede - ployments from humanitarian operations and an MSO exit...physi- cians from 15 nearby nations. Course graduates from several countries later deployed together to an earthquake response in Pisco, Peru , where

  17. Prevention of organic iodide formation in BWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Karjunen, T. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Laitinen, T.; Piippo, J.; Sirkiae, P. [VTT Manufacturing Technology (Finland)

    1996-12-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR`s as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs.

  18. Full system decontamination experience in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, N.; Sugai, K.; Katayouse, N.; Fujimori, A.; Iida, K.; Hayashi, K. [Tokyo Electric Power Company, Tokyo (Japan); Kanasaki, T.; Inami, I. [Toshiba Corporation, Yokohama (Japan); Strohmer, F. [Framatome ANP Gmbh, Eelangen (Germany)

    2002-07-01

    At the Fukushima Daiichi Nuclear Power Station unit 3, unit 2, unit 5 and unit 1 of Tokyo Electric Power Company (TEPCO), the replacement of the core shroud and internals has been conducted since 1997 in this order. The welded core internals in operating BWR plants were replaced to improve stress corrosion cracking (SCC) resistance. At present these units are operating smoothly. The developed technology concept is to restore those internals in open air inside the reactor pressure vessel (RPV). To reduce the radiation dose rate inside the RPV, not only a shielding method was applied to cut the radiation from the irradiated structures but also a chemical decontamination method was applied to dissolve the radioactive crud deposited on the surface by using chemical agents. The calculated decontamination factor (DF) at the RPV bottom reached 35-117. As result, the dose rate decreased to approximately 0.1 mSv/h under water. Before and after the installation of the in-vessel shielding, a mechanical cleaning was extensively applied inside the RPV to remove the residual crud as well as the swarf, chips from cutting. As a result, the dose rate at the RPV bottom decreased to ranging from 0.2 to 0.4 mSv/h in air. A working environment for human access, which was better than expected, was established inside the RPV, resulting in 70, 140, 50 and 70 man-Sv (estimated) saving respectively at unit 3 (1F-3), unit 2(1F-2), unit 5(1F-5) and unit 1(1F-1). All four full system decontamination (FSDs) contributed to the successful realization of the core shroud replacement project under the dry condition in RPV.

  19. System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

    2009-10-15

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  20. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  1. BWR online monitoring system based on noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Villafuerte, Javier [Departamento de Sistemas Nucleares, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, Ocoyoacac, Edo. de Mexico, 52750 (Mexico)]. E-mail: jov@nuclear.inin.mx; Castillo-Duran, Rogelio [Departamento de Sistemas Nucleares, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, Ocoyoacac, Edo. de Mexico, 52750 (Mexico)]. E-mail: rcd@nuclear.inin.mx; Alonso, Gustavo [Departamento de Sistemas Nucleares, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, Ocoyoacac, Edo. de Mexico, 52750 (Mexico)]. E-mail: galonso@nuclear.inin.mx; Calleros-Micheland, Gabriel [Central Nuclear de Laguna Verde, Comision Federal de Electricidad, Carr. Cardel-Nautla, km. 42.5, Alto Lucero, Veracruz (Mexico)]. E-mail: gcm9acpp@cfe.gob.mx

    2006-11-15

    A monitoring system for during operation early detection of an anomaly and/or faulty behavior of equipment and systems related to the dynamics of a boiling water reactor (BWR) has been developed. The monitoring system is based on the analysis of the 'noise' or fluctuations of a signal from a sensor or measurement device. An efficient prime factor algorithm to compute the fast Fourier transform allows the continuous, real-time comparison of the normalized power spectrum density function of the signal against previously stored reference patterns in a continuously evolving matrix. The monitoring system has been successfully tested offline. Four examples of the application of the monitoring system to the detection and diagnostic of faulty equipment behavior are presented in this work: the detection of two different events of partial blockage at the jet pump inlet nozzle, miss-calibration of a recirculation mass flow sensor, and detection of a faulty data acquisition card. The events occurred at the two BWR Units of the Laguna Verde Nuclear Power Plant. The monitoring system and its possible coupling to the data and processing information system of the Laguna Verde Nuclear Power Plant are described. The signal processing methodology is presented along with the introduction of the application of the evolutionary matrix concept for determining the base signature of reactor equipment or component and the detection of off normal operation conditions.

  2. Process inherent ultimate safety/boiling-water reactor PIUS/BWR

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.

    1985-01-01

    This document is a series of viewgraphs on: design basis of PIUS/BWR, definition of PIUS/BWR, mechanisms of safe shutdown and afterheat cooling, advantages of PIUS/BWR, and research and development requirements. (DLC)

  3. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  4. Pray for Peace. Prepare for War...and Stability Operations

    Science.gov (United States)

    2013-03-01

    ultimately achieve the national political objectives of the war.61 The Army’s experiences in Mexico were not an aberration, as the U.S. military...Report assessed that, even with the Army’s extensive experience with stability operations in “ Mexico , California, the Southern States, Cuba, …the...Manual: A Lasting Legacy in Today’s Counterinsurgency Warfare,” Historia 21 (2012): 54-62. 148 Gentile, “The Imperative for an American General

  5. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  6. Flex concept for US-A BWR extended loss of AC power events

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  7. BWR mechanics and materials technology update

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, E.

    1983-05-01

    This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration.

  8. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  9. New tools for the stability control; Nuevas herramientas para el control de la estabilidad

    Energy Technology Data Exchange (ETDEWEB)

    Melera, J.; Albendea, M.; Escriva, A.; Munoz-Cobos, L. L.; Montesinos

    2006-07-01

    IBERDROLA INGENIERIA y CONSTRUCCION, IBERDROLA GENERACION and the polytechnical Univisty of Valencia, are carrying out a project of R and D called DROP (Decay Ratio on-line Predictor) whose objective is to create a tool (predictor and monitor) for on-line monitoring of the thermal-hydraulic stability during the operation of BWR reactors. DROP is intended to be integrated in CAPRICOPE; the Core Monitoring System of Cofrentes NPP, to allow supervisors to known stability margins during the operation, simplifying reactor start-up and shutdown. In transient conditions, the knowledge of reactor stability will avoid reactor unplanned shutdowns. (Author) 11 refs.

  10. Conditional stability versus ill-posedness for operator equations with monotone operators in Hilbert space

    Science.gov (United States)

    Ioan Boţ, Radu; Hofmann, Bernd

    2016-12-01

    In the literature on singular perturbation (Lavrentiev regularization) for the stable approximate solution of operator equations with monotone operators in the Hilbert space the phenomena of conditional stability and local well-posedness or ill-posedness are rarely investigated. Our goal is to present some studies which try to bridge this gap. So we present new results on the impact of conditional stability on error estimates and convergence rates for the Lavrentiev regularization and distinguish for linear problems well-posedness and ill-posedness in a specific manner motivated by a saturation result. Taking into account that the behavior of the bias (regularization error in the noise-free case) is crucial, general convergence rates, including logarithmic rates, are derived for linear operator equations by means of the method of approximate source conditions. This allows us to extend well-known convergence rate results for the Lavrentiev regularization that were based on general source conditions to the case of non-selfadjoint linear monotone forward operators for which general source conditions fail. Examples presenting the self-adjoint multiplication operator as well as the non-selfadjoint fractional integral operator and Cesàro operator illustrate the theoretical results. Extensions to the nonlinear case under specific conditions on the nonlinearity structure complete the paper.

  11. BWR Refill-Reflood Program, Task 4. 7 - model development: TRAC-BWR component models

    Energy Technology Data Exchange (ETDEWEB)

    Cheung, Y K; Parameswaran, V; Shaug, J C

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation.

  12. Entanglement-assisted operator codeword stabilized quantum codes

    Science.gov (United States)

    Shin, Jeonghwan; Heo, Jun; Brun, Todd A.

    2016-05-01

    In this paper, we introduce a unified framework to construct entanglement-assisted quantum error-correcting codes (QECCs), including additive and nonadditive codes, based on the codeword stabilized (CWS) framework on subsystems. The CWS framework is a scheme to construct QECCs, including both additive and nonadditive codes, and gives a method to construct a QECC from a classical error-correcting code in standard form. Entangled pairs of qubits (ebits) can be used to improve capacity of quantum error correction. In addition, it gives a method to overcome the dual-containing constraint. Operator quantum error correction (OQEC) gives a general framework to construct QECCs. We construct OQEC codes with ebits based on the CWS framework. This new scheme, entanglement-assisted operator codeword stabilized (EAOCWS) quantum codes, is the most general framework we know of to construct both additive and nonadditive codes from classical error-correcting codes. We describe the formalism of our scheme, demonstrate the construction with examples, and give several EAOCWS codes

  13. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  14. Village Stability Operations in the Democratic Republic of Congo: A Special Operations Approach

    Science.gov (United States)

    2014-12-12

    of communication and logistics from Laos and Cambodia that extended into South Vietnam. Each team conducting CIDG operations dedicated assets and...then builds rapport with the villagers, develops their relationships with the elders, and begins to expand “white space;” the intangible span of... recognition of their important role in the overall conflict. A stability 97 Mushi, 16. 98 Rucogoza, 29. 99 United Nations Security Resolution 2147

  15. Development of a scatter search optimization algorithm for BWR fuel lattice design

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J.L.; Martin-del-Campo, C. [Mexico Univ. Nacional Autonoma, Facultad de Ingenieria (Mexico); Morales, L.B.; Palomera, M.A. [Mexico Univ. Nacional Autonoma, Instituto de Investigaciones en Matematicas Aplicadas y Sistemas, D.F. (Mexico)

    2005-07-01

    A basic Scatter Search (SS) method, applied to the optimization of radial enrichment and gadolinia distributions for BWR fuel lattices, is presented in this paper. Scatter search is considered as an evolutionary algorithm that constructs solutions by combining others. The goal of this methodology is to enable the implementation of solution procedures that can derive new solutions from combined elements. The main mechanism for combining solutions is such that a new solution is created from the strategic combination of two other solutions to explore the solutions' space. Results show that the Scatter Search method is an efficient optimization algorithm applied to the BWR design and optimization problem. Its main features are based on the use of heuristic rules since the beginning of the process, which allows directing the optimization process to the solution, and to use the diversity mechanism in the combination operator, which allows covering the search space in an efficient way. (authors)

  16. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  17. A New Methodology for Early Anomaly Detection of BWR Instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, K. N.

    2005-11-27

    The objective of the performed research is to develop an early anomaly detection methodology so as to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The research utilizes a model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, which is used as a generator of time series data for anomaly detection at an early stage. The model captures critical nonlinear features of coupled thermal-hydraulic and nuclear reactor dynamics and (slow time-scale) evolution of the anomalies as non-stationary parameters. The time series data derived from this nonlinear non-stationary model serves as the source of information for generating the symbolic dynamics for characterization of model parameter changes that quantitatively represent small anomalies. The major focus of the presented research activity was on developing and qualifying algorithms of pattern recognition for power instability based on anomaly detection from time series data, which later can be used to formulate real-time decision and control algorithms for suppression of power oscillations for a variety of anticipated operating conditions. The research being performed in the framework of this project is essential to make significant improvement in the capability of thermal instability analyses for enhancing safety, availability, and operational flexibility of currently operating and next generation BWRs.

  18. Characterization of corrosion layers on irradiated and non-irradiated surfaces in BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J.; Balek, V.; Zmitko, M.; Brozova, A.; Burda, J. [Nuclear Research Inst., Rez (Czech Republic); Hoffmann, H.; Ruehle, W. [VGB Essen (Germany); Bezdicka, P. [Institute of Inorganic Chemistry, ASCR, Rez (Czech Republic)

    2002-07-01

    Stress corrosion cracking of low-alloyed steel 22NiMoCr37 is evaluated with the goal to determine crack growth rate in irradiated steel under conditions simulating closely conditions of BWR RPV under operation. For the experiment, in pile BWR experimental loop has been built at Nuclear Research Institute, Rez. During the experiment, specimens are loaded by cyclic and constant load. Crack growth is monitored by means of potential drop measurement and COD. Corrosion layers formed on specimens in reactor water loop exposed to BWR primary water chemistry and radiation were studied. Two sets of specimens were placed in loop channels. One set of specimens was situated in reactor conditions and the second set out of reactor, other parameters like water chemistry (e.g. concentration of hydrogen, oxygen and conductivity), temperature and flow rate were identical. By means of this an effect of radiation could be studied. The differences in chemical composition, structure and microstructure of corrosion products were characterized by SEM and X-ray powder diffractometry. The differences in microstructure of corrosion layer formed under different conditions were observed. (authors)

  19. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  20. 14 CFR 25.672 - Stability augmentation and automatic and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... power-operated systems. 25.672 Section 25.672 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... Construction Control Systems § 25.672 Stability augmentation and automatic and power-operated systems. If the functioning of stability augmentation or other automatic or power-operated systems is necessary to...

  1. 14 CFR 29.672 - Stability augmentation, automatic, and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... power-operated systems. 29.672 Section 29.672 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... Construction Control Systems § 29.672 Stability augmentation, automatic, and power-operated systems. If the functioning of stability augmentation or other automatic or power-operated system is necessary to...

  2. 14 CFR 27.672 - Stability augmentation, automatic, and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... power-operated systems. 27.672 Section 27.672 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... Construction Control Systems § 27.672 Stability augmentation, automatic, and power-operated systems. If the functioning of stability augmentation or other automatic or power-operated systems is necessary to...

  3. Logical model for the control of a BWR turbine;Modelo logico para el control de una turbina de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R., E-mail: yonaeton@hotmail.co [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2009-07-01

    In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

  4. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde; Rango del monitor de radiacion para el venteo rigido de la contencion primaria durante operacion normal y emergencia para un reactor BWR-5 en Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  5. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  6. Stability and monotonicity of Lotka-Volterra type operators

    CERN Document Server

    Mukhamedov, Farrukh

    2009-01-01

    In the present paper, we study Lotka-Volterra (LV) type operators defined in finite dimensional simplex. We prove that any LV type operator is a surjection of the simplex. After, we introduce a new class of LV-type operators, called $M$LV type. We prove convergence of their trajectories and study certain its properties. Moreover, we show that such kind of operators have totaly different behavior than ${\\mathbf{f}}$-monotone LV type operators.

  7. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    Science.gov (United States)

    Ratnayake, Ruwan Kumara

    spacer grid locations. Portions of a prototypic 8 x 8 spacer grid and metal tubes simulating fuel rods were used for these experiments. Although the test section was unheated, the flow conditions were chosen to closely depict the normal operating conditions of a BWR core subchannel. Results confirmed that entrainment at spacer grid locations is significant. Subsequent COBRA-TF simulations mimicking enhanced entrainment and deposition at spacer grid locations showed marked effects on dryout prediction. This indicates that incorporation of a proper spacer grid model to COBRA-TF could in fact improve the overall code performance. A method of problem analysis by decoupling the overall spacer grid hydrodynamic phenomenon into several smaller individual mechanisms is used in this work. This document includes the initial conceptualization of entrainment and deposition mechanisms, their mathematical formulation and results obtained after integrating them to COBRA-TF.

  8. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  9. Analysis of BWR instabilities coupled with 3D code RELAP5 / PARCSv2.7. Application to the event happened in Oskarshamn-2 in 1999; Analisis de inestabilidades en BWR con el codigo acoplado 3D RELAP5/PARCSv2.7. Aplicacion al evento sucedido en Oskarshamn-2 en1999

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fenoll, M.; Barrachina, T.; Miro, R.; Verdu, G.

    2014-07-01

    In this work, part of our works in the frame of the OECD/NEA Oskarshamn-2 (O{sub 2}) BWR Stability Benchmark for Coupled Code Calculations and Uncertainty Analysis in Modelling are shown. The objective is to simulate the instability event registered in February 1999 at the Swedish NPP Oskarshamn-2 with the coupled code RELAP5/PARCSv2.7. (Author)

  10. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Gonzalez, Jesus S. [Univ. Politecnica de Madrid (Spain); Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William BJ J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  11. Measures of Effective Military Public Health Interventions in Stability Operations

    Science.gov (United States)

    2011-06-10

    March 2006), http://www.csrs-nps.org/ logistica /public/docs/game_report.pdf (Accessed 14 May 2011), 69. 39Pruden, 14. 40Himmler, 12. 41JP 4-02, I...Stabilization & Reconstruction Studies, Naval Postgraduate School, March 2006. http://www.csrs-nps.org/ logistica /public/docs/game_report.pdf (accessed

  12. Estimate of coolant flow in assemblies of a natural circulation BWR applying and equivalent electric model; Estimacion del flujo de refrigerante en los ensambles de un BWR de circulacion natural aplicando un modelo electrico equivalente

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: julfi_jg@yahoo.com.mx

    2009-10-15

    The present work exposes the design and implementation of an advanced controller that it allows to estimate the coolant flow in fuel assemblies of a natural circulation BWR in real time. the complete development of this study is part of a doctoral project in course. In this work the construction of optimal controller is shown that allows to estimate the coolant flows in reactor and its operation applied to an equivalent electric model to natural circulation ESBWR. The controller design that allows the completely automatic starter of natural circulation reactor, required of a variables estimator not meter directly of nuclear power plant and use of local distributions estimates of coolant flow, (this controller type at the moment is utilized in the A BWR and several BWR in operation in Japan). The construction of estimator controller is mathematically based in the theory referring to Kalman filter, whose algorithm provides an advanced control of system. To prove the estimator operation was developed a simplified model that reproduces the basic dynamic of coolant flowing in the ESBWR, a practice way and very interesting of representing this phenomenon is by means the use of an equivalent electric model, which was developed starting from analogies that there is among the relation that keep the pressure differences with the mass flow and differences of electric potential with electric current. A detailed analysis of equivalence among models will be presented in a later article. (Author)

  13. Study of transient turbine shot without bypass in a BWR; Estudio del transitorio disparo de turbina sin bypass en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  14. Study of transient rod extraction failure without RBM in a BWR; Estudio del transitorio error de extraccion de barra sin RBM en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

  15. Robust Stability and H∞ Stabilization of Switched Systems with Time-Varying Delays Using Delta Operator Approach

    Directory of Open Access Journals (Sweden)

    Chen Qin

    2013-01-01

    Full Text Available This paper considers the problems of the robust stability and robust H∞ controller design for time-varying delay switched systems using delta operator approach. Based on the average dwell time approach and delta operator theory, a sufficient condition of the robust exponential stability is presented by choosing an appropriate Lyapunov-Krasovskii functional candidate. Then, a state feedback controller is designed such that the resulting closed-loop system is exponentially stable with a guaranteed H∞ performance. The obtained results are formulated in the form of linear matrix inequalities (LMIs. Finally, a numerical example is provided to explicitly illustrate the feasibility and effectiveness of the proposed method.

  16. A stability analysis of a real space split operator method for the Klein-Gordon equation

    CERN Document Server

    Blumenthal, Frederick

    2011-01-01

    We carry out a stability analysis for the real space split operator method for the propagation of the time-dependent Klein-Gordon equation that has been proposed in J. Comput. Phys. 228 (24) (2009) 9092-9106. The region of algebraic stability is determined analytically by means of a von-Neumann stability analysis for systems with homogeneous scalar and vector potentials. Algebraic stability implies convergence of the real space split operator method for smooth absolutely integrable initial conditions. In the limit of small spatial grid spacings h in each of the d spatial dimensions and small temporal steps, the stability condition becomes h/{\\tau}>\\surddc for second order finite differences and \\surd3h/(2{\\tau})>\\surddc for fourth order finite differences, respectively, with c denoting the speed of light. Furthermore, we demonstrate numerically that the stability region for systems with inhomogeneous potentials coincides almost with the region of algebraic stability for homogeneous potentials.

  17. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  18. First Among Equals: Interagency Cooperation in Stability Operations

    Science.gov (United States)

    2016-05-26

    has a responsibility to “plan, conduct, and 1 Field Manual (FM) 3-07, Stability (Washington, DC: Government Printing Office, 2014), 3-18-3-19. 2...finding is an articulated line of authority running from the 36 Stephen Van Evera, Guide to Methods for Students of Political Science (Ithaca: Cornell...Aid to Greece” (Athens, Greece, June 15, 1948), 2, 22. 57 John O. Iatrides, Ambassador MacVeagh Reports: Greece, 1933-1947 ( Princeton , NJ: Princeton

  19. The Road Not Taken: Addressing Corruption During Stability Operations

    Science.gov (United States)

    2015-06-01

    Your red marks made us better students and writers. We would like to thank Professor George Lober for his generous feedback and sound advice . Your...after Daoud’s departure and signed a new constitution establishing a bicameral parliamentary system and elections. This new system of government...parties to use the parliamentary system against the royal family. Popular discontent further weakened Afghanistan’s stability. In 1972, student-led

  20. Numerical simulation of boron injection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

    2010-02-15

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

  1. 14 CFR 23.672 - Stability augmentation and automatic and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... power-operated systems. 23.672 Section 23.672 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... and power-operated systems. If the functioning of stability augmentation or other automatic or power-operated systems is necessary to show compliance with the flight characteristics requirements of this...

  2. An overview of the BWR ECCS strainer blockage issues

    Energy Technology Data Exchange (ETDEWEB)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-03-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, {open_quotes}Containment Emergency Sump Performance,{close_quotes} and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts.

  3. VIPRE-W / MEFISTO-T - A mechanistic tool for transient prediction of dryout in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, C., E-mail: carl.adamsson@psi.ch [Westinhouse Electric Sweden, Vasteras (Sweden); Paul Scherrer Institut, Villigen (Switzerland); Le Corre, J-M., E-mail: lecorrjm@westinghouse.com [Westinhouse Electric Sweden, Vasteras (Sweden)

    2011-07-01

    The VIPRE-W/MEFISTO-T code package constitutes a simplified approach to sub-channel film-flow analysis whereby the transport equations for the liquid films are decoupled from each other. The approach allows fast and robust simulation with high axial resolution of realistic BWR transients. It has previously been shown that a steady-state version of the model agrees well with dryout measurements in full-scale fuel assembly mock-ups performed at the Westinghouse FRIGG loop. In this paper, we present validation of the transient version of the code with around 300 transient dryout experiments from the same loop. The transients involve realistic variations of flow and power and three different axial power distributions at conditions typical for BWR operation. The results from the film-flow analysis show high precision in the dryout prediction but a hitherto unexplained bias that reduces the accuracy. (author)

  4. Stability of the inverse resonance problem for Jacobi operators

    CERN Document Server

    Bledsoe, Matthew

    2012-01-01

    When the coefficients of a Jacobi operator are finitely supported perturbations of the 1 and 0 sequences, respectively, the left reflection coefficient is a rational function whose poles inside, respectively outside, the unit disk correspond to eigenvalues and resonances. By including the zeros of the reflection coefficient, we have a set of data that determines the Jacobi coefficients up to a translation as long as there is at most one half-bound state. We prove that the coefficients of two Jacobi operators are pointwise close assuming that the zeros and poles of their left reflection coefficients are $\\eps$-close in some disk centered at the origin.

  5. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  6. Stability Analysis for Operation of DG Units in Smart Grids

    DEFF Research Database (Denmark)

    Pouresmaeil, Edris; Shaker, Hamid Reza; Mehrasa, Majid

    2015-01-01

    This paper presents a multifunction control strategy for the stable operation of Distributed Generation (DG) units during grid integration. The proposed control model is based on Direct Lyapunov Control (DLC) theory and provides a stable region for the appropriate operation of DG units during grid...... integration. Using DLC technique in DG technology can provide the continuous injection of maximum active power in fundamental frequency from the DG source to the grid, compensating all reactive power and harmonic current components of grid-connected loads through the integration of DG link into the grid....... Application of this concept can guarantee to reduce the stress on the grid during the energy demand peak. Simulation results are presented to demonstrate the proficiency and performance of the proposed DLC technique in DG technology....

  7. Stability Operations in Somalia 1992-1993: A Case Study

    Science.gov (United States)

    2016-07-01

    particularly in the north. These groups included Islamic Unity, which op- posed the SSDF in the north, and the Muslim Brotherhood, which operated...Organization of the Islamic Conference (OIC), began an initiative to restore peace in Somalia. His term as Secretary-General expired on December 31...13th Spt Comd) JSOFOR AFFOR (437th Wing) JPOTF MARFOR (I MEF) 34 General, Major General Steven L. Arnold ( USA ), knew General Johnston well

  8. Detainee Health Care: Essential Element of Stability Operations

    Science.gov (United States)

    2012-03-22

    food , by mutilation, death, or any other barbarity. 31 Missing from the 1864 and 1906 Geneva Conventions were guidelines for the treatment of able...government.74 Accordingly the host nation Ministry of Health ( MoH ) must eventually assume the health care of detainees. This requires a functioning health...care system and alignment of current detainee medical operations with the MoH capability without reducing the standard of care. This process begins

  9. Village Stability Operations and the Afghan Local Police

    Science.gov (United States)

    2014-10-01

    professionals and the Special Operations Forces (SOF) students and leaders for consideration by the SOF community and defense leadership. JSOU is the...one of the most acclaimed and controversial histories of the past decade, having been the subject of an academic confer- ence at Williams College ...expropriated the lands of Pashtuns and used the police to brutalize those who dared resist him. Mistreatment of Pashtuns for ethnic reasons became a

  10. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  11. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  12. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  13. Landsat-8 Operational Land Imager Radiometric Calibration and Stability

    Directory of Open Access Journals (Sweden)

    Brian Markham

    2014-12-01

    Full Text Available The Landsat-8 Operational Land Imager (OLI was radiometrically calibrated prior to launch in terms of spectral radiance, using an integrating sphere source traceable to National Institute of Standards and Technology (NIST standards of spectral irradiance. It was calibrated on-orbit in terms of reflectance using diffusers characterized prior to launch using NIST traceable standards. The radiance calibration was performed with an uncertainty of ~3%; the reflectance calibration to an uncertainty of ~2%. On-orbit, multiple calibration techniques indicate that the sensor has been stable to better than 0.3% to date, with the exception of the shortest wavelength band, which has degraded about 1.0%. A transfer to orbit experiment conducted using the OLI’s heliostat-illuminated diffuser suggests that some bands increased in sensitivity on transition to orbit by as much as 5%, with an uncertainty of ~2.5%. On-orbit comparisons to other instruments and vicarious calibration techniques show the radiance (without a transfer to orbit adjustment, and reflectance calibrations generally agree with other instruments and ground measurements to within the uncertainties. Calibration coefficients are provided with the data products to convert to either radiance or reflectance units.

  14. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

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

  15. Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

    2009-10-15

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  16. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cheng L. Y.; Baek J.; Cuadra,A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  17. Analysis of containment venting following a core damage at a BWR Mark I using THALES-2

    Energy Technology Data Exchange (ETDEWEB)

    Widodo, Surip [Nuclear Safety Technology Development Center, National Nuclear Energy Agency (BATAN), Tangerang (Indonesia); Ishikawa, Jun; Muramatsu, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Sakamoto, Toru [Toshiba Advanced System Co., Kawasaki, Kanagawa (Japan)

    2000-11-01

    Analysis of containment venting following a core damage at a boiling water reactor (BWR) Mark I using THALES-2 was performed. In this analysis, the effect of various parameters, namely, the areas of the vent path, containment venting pressure, and accident sequences on the containment thermodynamic response, and radionuclide transport and release in the containment venting at a BWR was examined. The code THALES-2B developed by the Japan Atomic Energy Research Institute (JAERI) was used in this analysis. The model plant in this analysis was the Browns Ferry plant. From this analysis was found that the 4-inch pipe of containment venting flow path is sufficient to maintain the containment pressure in the specified range if the containment was pressurized by the decay heat power. The entrainment by the pool swelling as well as by the flashing was not occurred during the containment venting. The source terms are not sensitive to the variation of containment venting flow path area. The containment venting pressure operation setting point has important rule in the containment venting. In the containment venting, the source terms are not sensitive to the accident sequence, except for Sr source term. In order to get better understanding on the containment venting strategy, the following analyses are necessary. Analyses of accident sequence which has a high power such as anticipated transient without scram are necessary, as well as analyses of accident sequence which pressurize the containment before the core damage. (author)

  18. Estimation of dose rate around the spent control rods of a BWR; Estimacion de la rapidez de dosis alrededor de las barras de control gastadas de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cancino P, G.

    2016-10-01

    The energy can come from fossil renewable sources (solar (natural gas, oil), wind, hydro, tidal, geothermal, biomass, bio energy and nuclear. Nuclear power can be obtained by fission reactions and fusion (still under investigation) atomic nuclei. Fission, is a partition of a very heavy nucleus (Uranium 235, for example) into two lighter nuclei. Much of the world's electric power is generated from the energy released by fission processes. In a nuclear power reactor, light water as the BWR, there are many important elements that allow safe driving operation, one of them are the elements or control systems, the burnable poison or neutron absorber inherently allow control power reactor. The control rods, which consist mostly of stainless steel and absorbing elements (such as boron carbide, hafnium, cadmium, among others) of thermal neutrons is able to initiate, regulate or stop the reactor power. These, due to the use of depleted burned or absorbing material and therefore reach their lifespan, which can be 15 years or have other values depending on the manufacturer. Control rods worn should be removed, stored or confined in expressly places. Precisely at this stage arises the importance of knowing their radiological condition to manipulate safely and without incident to the people health responsible for conducting these proceedings state arises. This thesis consists in the estimation of the dose rate in spent control rod made of boron carbide, from a typical BWR reactor. It will be estimated by direct radiation measurements with measurement equipment for radiotherapy ionization chamber, in six spent control rods, which were taken at different reactor operating cycles and are in a spent fuel pool. Using bracket electromechanical and electronic equipment for positioning and lifting equipment for radiation measurement around the control rod in the axial and radial arrangement for proper scanning. Finally will be presented a graphic corresponding to the dose

  19. A stability analysis of a real space split operator method for the Klein-Gordon equation

    Science.gov (United States)

    Blumenthal, Frederick; Bauke, Heiko

    2012-01-01

    We carry out a stability analysis for the real space split operator method for the propagation of the time-dependent Klein-Gordon equation that has been proposed in Ruf et al. [M. Ruf, H. Bauke, C.H. Keitel, A real space split operator method for the Klein-Gordon equation, Journal of Computational Physics 228 (24) (2009) 9092-9106, doi:10.1016/j.jcp.2009.09.012]. The region of algebraic stability is determined analytically by means of a von-Neumann stability analysis for systems with homogeneous scalar and vector potentials. Algebraic stability implies convergence of the real space split operator method for smooth absolutely integrable initial conditions. In the limit of small spatial grid spacings h in each of the d spatial dimensions and small temporal steps τ, the stability condition becomes h/τ>√{d}c for second order finite differences and √{3}h/(2τ)>√{d}c for fourth order finite differences, respectively, with c denoting the speed of light. Furthermore, we demonstrate numerically that the stability region for systems with inhomogeneous potentials coincides almost with the region of algebraic stability for homogeneous potentials.

  20. Evaluation of nickel-yttria stabilized zirconia anode degradation during discharge operation and redox cycles operation by electrochemical calculation

    Science.gov (United States)

    Shimura, Takaaki; Jiao, Zhenjun; Shikazono, Naoki

    2016-10-01

    Degradation of Solid Oxide Fuel Cell (SOFC) anode during discharge operation and redox cycles operation were evaluated by three-dimensional electrochemical calculations using a Lattice Boltzmann method (LBM). Three dimensional microstructures were obtained by Focused Ion Beam Scanning Electron Microscopy (FIB-SEM) reconstruction. In the electrochemical calculations, changes in exchange current density and ionic conductivity of Yttria stabilized Zirconia (YSZ) during the operations were assumed and their values were calculated by fitting the calculated overpotential values to the experimental ones. Changes in triple phase boundary density calculated from the reconstructed microstructures were inconsistent with the gradual degradation observed during repeated redox-discharge cycles. Changes of the fitted exchange current density and YSZ ionic conductivity values in both discharge operation and redox cycle operation showed same tendency as the experimental results. Change in exchange current density or YSZ ionic conductivity should be considered as an essential factor which governs the cell performance change regardless of the redox treatment.

  1. Virasoro frames and their Stabilizers for the E_8 lattice type Vertex Operator Algebra

    OpenAIRE

    2001-01-01

    The concept of a framed vertex operator algebra was studied in [DGH] (q-alg/9707008). This article is an analysis of all Virasoro frame stabilizers of the lattice VOA V for the E_8 root lattice, which is isomorphic to the E_8-level 1 affine Kac-Moody VOA V. We analyze the frame stabilizers, both as abstract groups and as subgroups of the Lie group Aut(V) = E_8(C). Each frame stabilizer is a finite group, contained in the normalizer of a 2B-pure elementary abelian 2-group in Aut(V). In particu...

  2. BWR refill-reflood program: core spray distribution experimental task plan

    Energy Technology Data Exchange (ETDEWEB)

    Eckert, T.

    1981-02-01

    An experimental task plan for the BWR/4 core spray task of the Refill-Reflood Test Program is presented. The test program will provide core spray distribution data for a 30 degree sector of the BWR/4 and 5-218 design. This design uses different nozzle types and different sparger elevations than the BWR/6-218 design which was tested previously. Test parameter ranges are specified; individual tests are defined; and measurement and data utilization plans are defined.

  3. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  4. Power System Stabilizer Based on Robust H∞ Controller for Low Frequency Operating Range

    Directory of Open Access Journals (Sweden)

    Ali Mohamed Yousef

    2012-05-01

    Full Text Available The aim of study is designed of Power System Stabilizer (PSS based on H4 approach for power system stabilization. The uncertainties in power system modeling and operations are considered at designing of H∞ PSS. The bounds of power system parameters are determined over a wide range of low frequency operating conditions. These bounds are used to design a robust H∞ PSS. A sample power system composed a synchronous generator connected to infinite bus through transmission line is simulated. The digital H∞ PSS can achieve good performance over a wide range of operating conditions. A comparison between power system responses at variety of operating conditions using the proposed H∞ PSS and Linear Quadratic Regulator LQR control have been done. H2 PSS is designed and compared with the proposed controller.

  5. Technical Specification action statements requiring shutdown. A risk perspective with application to the RHR/SSW systems of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy, Espoo (Finland); Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States)

    1993-11-01

    When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements.

  6. Mission Command or Mission Failure?: Evaluation Decentralized Execution in Contemporary Stability Operations

    Science.gov (United States)

    2014-05-15

    operations in OEF/OIF are attributable not to underlying weaknesses in mission command as a theoretical construct, or its lack of suitability for...posited that acknowledged shortcomings in the success of stability operations in OEF/OIF are attributable not to underlying weaknesses in mission command...230. 7 Jeffrey Buchman, Maxie Y. Davis, and Lee T. Wright. “Death of the Combatant Command? Toward a Joint Interagency Approach.” Joint Force

  7. Two-staged operation on C2 neoplastic lesions: anterior excision and posterior stabilization.

    Science.gov (United States)

    Colak, Ahmet; Kutlay, Murat; Kibici, Kenan; Demircan, M Nusret; Akin, Osman N

    2004-07-01

    This retrospective study included eight consecutive cases with C2 vertebral body neoplastic lesions. The anterior retropharyngeal approach was used to remove the lesions and decompress the spinal cord. Spinal stabilization with occipitocervical plating in a second-stage operation makes the treatment more tolerable for patients. The object of this study was to determine the effectiveness of a two-stage operation strategy for these lesions. Eight patients were operated on via anterior retropharyngeal approach and then stabilized with occipitocervical plates posteriorly in a second sitting. All neck pain and all dysphagia problems resolved. Partial neurologic improvement was achieved in three out of four patients. No postoperative infection was seen. The retropharyngeal approach to the upper cervical spine and anterior foramen magnum lesions is an effective alternative to transoral surgery because of low complication rates. Neoplastic lesions in the upper cervical spine can safely and effectively be operated with this technique. The general medical status of patients with malignancies does not permit too long, time-consuming operations. Stabilization of the spine in a separate operation increases patient tolerability without any morbidity.

  8. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  9. Prosperity or Perdition: Do Lines of Operations Apply in Stability Operations?

    Science.gov (United States)

    2003-01-01

    explains and defines lines of operations in familiar linear terms, specifying geographic, interior, and exterior examples. It also proposes a new...groups loyal to the two parties extended the blood feud into the rural areas of Colombia. During this fifteen-year period, known as La Violencia (The...the later states of La Violencia , a disillusioned, largely rural, minority movement emerged to form “Peasant Republics” in the country-side not

  10. An approximation theory or strongly stabilizing solutions to the operator LQ Riccati equation

    NARCIS (Netherlands)

    Oostveen, JC; Curtain, RF; Ito, K

    2000-01-01

    The linear-quadratic (LQ) control problem is considered for a class of infinite-dimensional systems with bounded input and output operators, that are not exponentially stabilizable, but only strongly stabilizable. A sufficient condition for the existence of a minimizing control and of a stabilizing

  11. Recovery of Metals and Stabilization of Arsenic from (Bio-)Leaching Operations by Engineered Biological Processes

    NARCIS (Netherlands)

    Gonzalez Contreras, P.A.; weghuis, M.O.; Weijma, J.; Buisman, C.J.N.

    2013-01-01

    This paper focuses on the application of biotechnological stabilization of arsenic from (bio-) leaching operations. One of the latest applications of the Thioteq technology is arsenic immobilization. The Thioteq-scorodite biorecovery reactor is an aerobic system to immobilise arsenic in bio-scorodit

  12. CVD Diamond Detector Stability Issues for Operation at the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Schmid, G J; Koch, J A; Moran, M J; Lerche, R A; Izumi, N; Phillips, T W; Glebov, V Y; Sangster, T C; Stoeckl, C

    2003-08-22

    Synthetic diamond crystals produced by the Chemical Vapor Deposition (CVD) technique can serve as fast, radiation hard, neutron sensors for the National Ignition Facility (NIF). Here we explore the stability issues, such as charge trapping and high-flux saturation, that will be relevant to operation at the NIF.

  13. Operational electrochemical stability of thiophene-thiazole copolymers probed by resonant Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Wade, Jessica; Wood, Sebastian; Kim, Ji-Seon, E-mail: ji-seon.kim@imperial.ac.uk [Department of Physics and Centre for Plastic Electronics, Imperial College London, London SW7 2AZ (United Kingdom); Beatrup, Daniel; Hurhangee, Michael; McCulloch, Iain; Durrant, James R. [Department of Chemistry and Centre for Plastic Electronics, Imperial College London, London SW7 2AY (United Kingdom); Bronstein, Hugo [Department of Chemistry and Centre for Plastic Electronics, Imperial College London, London SW7 2AY (United Kingdom); Department of Chemistry, University College London, London WC1H 0AJ (United Kingdom)

    2015-06-28

    We report on the electrochemical stability of hole polarons in three conjugated polymers probed by resonant Raman spectroscopy. The materials considered are all isostructural to poly(3-hexyl)thiophene, where thiazole units have been included to systematically deepen the energy level of the highest occupied molecular orbital (HOMO). We demonstrate that increasing the thiazole content planarizes the main conjugated backbone of the polymer and improves the electrochemical stability in the ground state. However, these more planar thiazole containing polymers are increasingly susceptible to electrochemical degradation in the polaronic excited state. We identify the degradation mechanism, which targets the C=N bond in the thiazole units and results in disruption of the main polymer backbone conjugation. The introduction of thiazole units to deepen the HOMO energy level and increase the conjugated backbone planarity can be beneficial for the performance of certain optoelectronic devices, but the reduced electrochemical stability of the hole polaron may compromise their operational stability.

  14. Stability analysis of slot-entry hybrid journal bearings operating with non-newtonian lubricant

    Directory of Open Access Journals (Sweden)

    H.C. Garg

    2015-09-01

    Full Text Available This paper presents theoretical investigations of rheological effects of lubricant on stability parameters of various configurations of slot-entry hybrid journal bearing system. FEM has been used to solve Reynolds equation governing flow of lubricant in bearing clearance space along with restrictor flow equation using suitable iterative technique. The non-Newtonian lubricant has been assumed to follow cubic shear stress law. The stability parameters in terms of stiffness coefficients, damping coefficients, threshold speed and whirl frequency of different configurations of slot-entry hybrid journal bearing have been computed and presented for wide range of external load while operating with Newtonian and Non-Newtonian lubricants. The computed results reveal that variation of viscosity due to non-Newtonian behavior of lubricant affects bearing stability quite significantly. The results are presented in graphical form and logical conclusions are drawn to identify best possible configuration from stability point of view.

  15. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  16. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  17. BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

    Energy Technology Data Exchange (ETDEWEB)

    Sicard, D.; Verdier, A. [COGEMA Logistics (AREVA Group) (France); Monsigny, P.A. [NOK/KKL (Switzerland)

    2004-07-01

    The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators.

  18. Airpower in contemporary interventions: stabilization, coercion, peace enforcement and peacemaking. Case: Operation Unified Protector, Libya 2011

    Directory of Open Access Journals (Sweden)

    Denisa SAVA

    2013-06-01

    Full Text Available This paper intends to analyze whether airpower is decisive and irreplaceable in putting an end to violent conflict nowadays, evaluating airpower as an instrument in contemporary warfare and its role in peace operations. The research will seek to confirm or invalidate the hypothesis according to which airpower can successfully take the lead in the Western Allies’ contemporary stabilization operations given that it reduces costs in terms of human lives and resources. Moving forward, the research will focus on airpower characteristics in the Libyan case, with an analysis of NATO’s Operation Unified Protector.

  19. Stabilization and reconstruction operations: the role of the US Army Veterinary Corps.

    Science.gov (United States)

    Smith, John C

    2007-01-01

    Stabilization and reconstruction operations in failed or failing states are vital to US security interests. These operations require a bottom-up approach, focusing on the population as the strategic center of gravity. This bottom-up approach must address the population's basic needs, as defined by Dr Abraham Maslow's hierarchy of needs, and provide a long-term means of self-sufficiency, rather than creating an "aid dependent economy." Focusing operations on agricultural projects provides relief from donor dependency, stimulates economic growth, and thwarts the power of spoilers. US Army Veterinary Corps personnel provide essential services ensuring the procurement of safe and wholesome subsistence and provision of medical care to government-owned animals. Veterinary Corps officers are also uniquely qualified to design and implement agricultural stabilization and reconstruction programs in conjunction with host-state ministries and agencies across the full range of military operations. Early, sustained engagement by veterinarians stimulates agricultural productivity, improves animal and human health, directly supports the population's hierarchy of needs on all levels, and accelerates stabilization operations by reducing the population's susceptibility to spoilers.

  20. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  1. A qualitative content analysis of global health engagements in Peacekeeping and Stability Operations Institute's stability operations lessons learned and information management system.

    Science.gov (United States)

    Nang, Roberto N; Monahan, Felicia; Diehl, Glendon B; French, Daniel

    2015-04-01

    Many institutions collect reports in databases to make important lessons-learned available to their members. The Uniformed Services University of the Health Sciences collaborated with the Peacekeeping and Stability Operations Institute to conduct a descriptive and qualitative analysis of global health engagements (GHEs) contained in the Stability Operations Lessons Learned and Information Management System (SOLLIMS). This study used a summative qualitative content analysis approach involving six steps: (1) a comprehensive search; (2) two-stage reading and screening process to identify first-hand, health-related records; (3) qualitative and quantitative data analysis using MAXQDA, a software program; (4) a word cloud to illustrate word frequencies and interrelationships; (5) coding of individual themes and validation of the coding scheme; and (6) identification of relationships in the data and overarching lessons-learned. The individual codes with the most number of text segments coded included: planning, personnel, interorganizational coordination, communication/information sharing, and resources/supplies. When compared to the Department of Defense's (DoD's) evolving GHE principles and capabilities, the SOLLIMS coding scheme appeared to align well with the list of GHE capabilities developed by the Department of Defense Global Health Working Group. The results of this study will inform practitioners of global health and encourage additional qualitative analysis of other lessons-learned databases.

  2. Optimized clearing work concept for the BWR containment; Optimiertes Raeumungskonzept fuer SWR-Sicherheitsbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Kraps, Uwe [AREVA NP GmbH (Germany)

    2012-11-01

    Based on the experiences of reactor dismantling in the NPPs Wuergasse, Obrigheim and Stade an optimized clearing work concept for the BWR containment including the reactor pressure vessel and the biological shield was developed. The concept is focused on the safety objective, the reduction of the collective dose and the reduction of the execution time. Precondition for the decommissioning license was up to now the removal of fuel elements from the reactor; due to the significantly increased period until fulfillment of this premises concepts are developed that can be performed with simultaneous reduction of the radiological inventories and the fire loads. The most important step of the guideline of the concept is the transition from hot to cold. The in-situ disassembling of the reactor internals can be performed with decreased water level in the reactor pressure vessel, with following water treatment and complete shutdown of operational systems. This status allows an accelerated further dismantling of the plant.

  3. Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

    Energy Technology Data Exchange (ETDEWEB)

    Su, T.M.

    1982-10-01

    Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report.

  4. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  5. Cofrentes EOC16B poolside measurements of channels from the three BWR vendors

    Energy Technology Data Exchange (ETDEWEB)

    Sedano, Pablo J. Garcia; Ayuela, Javier Iglesias [Iberdrola Ingenieria Construccion SAU, Veronica Anaures (Mexico); Albendea, Manuel [Iberdrola Generation S. A., Plaza Euskadi, 5 48009 BILBAO (Spain)

    2008-10-15

    As part of the EPRI Fuel Reliability Program, a fuel channel focus group was formed in 2002 to initiate measurements on irradiated BWR fuel channels. Fuel channels from GNF and AREVA have been measured in campaigns performed during 2004{approx}2007. Fuel channels designed and supplied by Westinghouse were of particular interest since no measurement information had been previously taken on modern Westinghouse channels operating on conventional loading pattern cycles, either in European or U.S. plants. Conventional loading pattern cycles are more susceptible to experience shadow corrosion induced bow since the fresh bundles are exposed to control blade influence early in life. During summer of 2007 extensive poolside measurements of a total of 180 fuel channels (24 SVEA-96 +/L, 68 SVEA-96 Optima-2, 36 GE-12, 42 GE-14 and 10 ATRIUM-10XP) have been performed by Westinghouse at Cofrentes NPP (Spanish BWR-6 operating on 24 month cycle strategy). This campaign has been co-sponsored by EPRI, Iberdrola and Westinghouse Sweden. Channels covering a range of exposure and control blade history were selected in order to determine the dependency of the channel deformation with those parameters. Channels with the most limiting conditions of exposure and control blade history were included. Channel bow, bulge and twist have been measured and fast neutron fluence calculations have been performed in order to determine the effects of neutron fluence gradient and shadow corrosion on the total channel deformation. Additionally channel oxide measurements have been performed on 20 channels from the three fuel vendors.The results indicate that channel bow and bulge remained at anticipated levels with no indication of significant channel bow due to shadow corrosion phenomenon. Destructive metallographic evaluations of samples taken from one cycle Westinghouse channels with high control blade exposure are underway at Studsvik hot cell facilities. These examinations will provide additional

  6. Automatic approach to stabilization and control for multi robot teams by multilayer network operator

    Directory of Open Access Journals (Sweden)

    Diveev Askhat

    2016-01-01

    Full Text Available The paper describes a novel methodology for synthesis a high-level control of autonomous multi robot teams. The approach is based on multilayer network operator method that belongs to a symbolic regression class. Synthesis is accomplished in three steps: stabilizing robots about some given position in a state space, finding optimal trajectories of robots’ motion as sets of stabilizing points and then approximating all the points of optimal trajectories by some multi-dimensional function of state variables. The feasibility and effectiveness of the proposed approach is verified on simulations of the task of control synthesis for three mobile robots parking in the constrained space.

  7. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors; Tendencias recientes en la mitigacion del IGSCC mediante modificaciones en la quimica del agua de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Diaz S, A.; Robles, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

  8. Improve the operational stability of the inverted organic solar cells using bilayer metal oxide structure.

    Science.gov (United States)

    Chang, Jingjing; Lin, Zhenhua; Jiang, Changyun; Zhang, Jie; Zhu, Chunxiang; Wu, Jishan

    2014-11-12

    Operational stability is a big obstacle for the application of inverted organic solar cells (OSCs), however, less talked about in the research reports. Due to photoinduced degradation of the metal oxide interlayer, which can cause shunts generation and degeneration in ZnO interlayer, a significant degradation of open circuit voltage (Voc) and fill factor (FF) has been observed by in situ periodic measurements of the device current density-voltage (J-V) curves with light illumination. By combining TiOx and ZnO to form bilayer structures on ITO, the photovoltaic performance is improved and the photoinduced degradation is reduced. It was found that the device based on ZnO/TiOx bilayer structure achieved better operational stability as compared to that with ZnO or TiOx interlayer.

  9. Special plate stability problems encountered in designing steel frameworks supporting large machinery operated in opencast mines

    Energy Technology Data Exchange (ETDEWEB)

    Ambrozy, G.

    1986-01-01

    The construction of large machinery that operates in opencast mines should by preceded by stability calculations as accurately as possible. The mathematical solution of the task is described. The method involves the determination of the coefficient matrix and of the coordinate functions. The accuracy of the results is estimated. Nomograms of the results obtained by using a computer program are presented. A numeric example verifies the validity of the mathematical solution.

  10. MATHEMATICAL MODEL OF POWER CONSUMPTION FOR SOME OIL PIPE-LINE SECTIONS WITH POOR OPERATIONAL STABILITY

    Directory of Open Access Journals (Sweden)

    J. N. Kolesnik

    2005-01-01

    Full Text Available Mathematical model of power consumption for technologically completed and non-completed oil pipe-line sections with poor operational stability has been developed on the basis of daily indices concerning oil transportation regimes. The model permits to take into account tendencies in power consumption under various time prediction cycles and ranges of oil freight turnover, changes in the bulk and characteristics of the transported oil, configuration and design parameters of oil pipe-line.

  11. Operational Modal Analysis Based on Subspace Algorithm with an Improved Stabilization Diagram Method

    Directory of Open Access Journals (Sweden)

    Shiqiang Qin

    2016-01-01

    Full Text Available Subspace-based algorithms for operational modal analysis have been extensively studied in the past decades. In the postprocessing of subspace-based algorithms, the stabilization diagram is often used to determine modal parameters. In this paper, an improved stabilization diagram is proposed for stochastic subspace identification. Specifically, first, a model order selection method based on singular entropy theory is proposed. The singular entropy increment is calculated from nonzero singular values of the output covariance matrix. The corresponding model order can be selected when the variation of singular entropy increment approaches to zero. Then, the stabilization diagram with confidence intervals which is established using the uncertainty of modal parameter is presented. Finally, a simulation example of a four-story structure and a full-scale cable-stayed footbridge application is employed to illustrate the improved stabilization diagram method. The study demonstrates that the model order can be reasonably determined by the proposed method. The stabilization diagram with confidence intervals can effectively remove the spurious modes.

  12. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  13. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity. The feedback reactivity is a function of the operating conditions of the system, and cannot be directly measured. However, power measurements are relatively easy to obtain from the nuclear instrumentation and process computer, and are used in conjunction with a reduced order model to estimate the gain of the thermal-hydraulics feedback using an Extended Kalman Filter. The reduced order model is obtained by estimating the thermal-hydraulic transfer function from the frequency-domain BWR code LAPUR, and the stability properties are evaluated based on the pair of complex conjugate eigenvalues. Because of the recursive nature of the Kalman Filter, an estimate of the decay ratio is generated every sampling time, allowing continuous estimation of the stability parameters. A test platform based on a nuclear-coupled boiling channel is developed to validate the capability of the BWR stability monitoring methodology. The thermal-hydraulics for the boiling channel is modeled and coupled with neutron kinetics to analyze the non-linear dynamics of the closed-loop system. The model uses point kinetics to study core-wide oscillations, and normalized modal kinetics are introduced to study out-of-phase oscillations. The coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate noisy power time series. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions

  14. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  15. Using operational and defined fractions to assess soil organic matter stabilization and structure

    Science.gov (United States)

    Horwath, W. R.

    2015-12-01

    Studies on soil organic matter (SOM) began with alkaline solvents revealing a dark colored substance that could be isolated under low pH. Further studies revealed fulvic and humic acids and humin fractions leading to theories on functional groups and metal-clay bridging mechanisms. The fate of isotopes in these fractions revealed soil carbon pools with varying turnover rates with half the soil carbon (C) in humin and acid hydrolyzed fractions over 1000 years old. These results are the basis of the three pool conceptual framework used in many biogeochemical models. Theories on the role of functional groups and compound classes further elaborated concepts on physical (aggregates) and chemical mechanisms of C stabilization. With the advance of analytical instrumentation, the operational fractions were further defined to the compound and molecular levels. These studies confirmed the majority of soil C is microbially derived. Our observation that all microbial groups contributed nonselectively to soil C maintenance independent of mineralogy suggests that compound characteristics within integrated structures are more important than the source of individual compounds for stabilizing soil C. In dissolved organic C floccing studies using Near Edge X-ray Fine Structure analysis, we found that aromatic compounds interacted first with Fe, however, the majority of direct bonds to Fe were polysaccharides, reinforcing that an integrative chemical structure rather than direct bonds imparted stability in organo-metal interactions. Using a novel differential scanning calorimeter coupled to an isotope ratio mass spectrometer setup, we confirmed that the presence of clays (independent of clay type) increased the microbial utilization of calcium stabilized high versus low temperature compounds, asserting that higher temperature compounds (i.e., phenolics) are likely less tightly bound by clay minerals. The integration of operational and defined fractions of SOM remains a legitimate

  16. Development of a numerical code for the analysis of the linear stability of the U1 and U2 reactors of the CNLV; Desarrollo de un codigo numerico para el analisis de estabilidad lineal de los reactores de las U1 y U2 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Estrada P, C.E. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Nunez C, A.; Amador G, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Mexico D.F. (Mexico)

    2001-07-01

    The computer code ANESLI-1 developed by the CNSNS and UAM-I, has the main goal of making stability analysis of nuclear reactors of the BWR type, more specifically, the reactors of the U1 and U2 of the CNLV. However it can be used for another kind of applications. Its capacity of real time simulator, allows the prediction of operational transients, and conditions of dynamic steady states. ANESLI-1 was developed under a modular scheme, which allows to extend or/and to improve its scope. The lineal stability analysis predicts the instabilities produced by the wave density phenomenon. (Author)

  17. A study on threshold voltage stability of low operating voltage organic thin-film transistors

    Science.gov (United States)

    Padma, N.; Sen, Shaswati; Sawant, Shilpa N.; Tokas, R.

    2013-08-01

    A low operating voltage (<2 V) organic field-effect transistor (OFET) using phenylhexyltrichlorosilane (PTS) self-assembled monolayer (SAM) dielectric and copper phthalocyanine (CuPc) as semiconductor with improved mobility (0.035 cm2 V-1 s-1) and threshold voltage stability was demonstrated. This device showed better performance when compared to an OFET with octyltrichlorosilane (OTS-8) SAM dielectric. The improved mobility was attributed to the 2D growth mode of CuPc on PTS SAM because of surface energy matching between the two, whereas CuPc film on OTS-8 showed a 3D growth mode with larger grain boundary density. The higher threshold voltage stability of OFETs on PTS SAM was attributed to the efficient coverage and screening of trap centres at dielectric/semiconductor interface due to stronger intermolecular linking and formation of closely packed surface by the bulky phenyl end groups. Decrease in grain boundaries offered by 2D growth of CuPc for electron and hole trapping was also found to be another reason for improved threshold voltage stability. The results indicated that the nature of the end group of SAM dielectric, surface chemistry of dielectric and initial growth mode of semiconductors are all responsible for improvement in threshold voltage stability and enhanced performance of OFET.

  18. RF phase stability in the 100-MeV proton linac operation

    Science.gov (United States)

    Seol, Kyung-Tae

    2015-02-01

    The 100-MeV proton linac of the Korea multi-purpose accelerator complex (KOMAC) has been operated to provide a proton beam to users. The 100-MeV linac consists of a 3-MeV radio-frequency quadrupole accelerator (RFQ), four 20-MeV drift-tube linac (DTL) tanks, two medium-energy beam-transmitter (MEBT) tanks, and seven 100-MeV DTL tanks. The requirements of the field stability are within ±1% in RF amplitude and ±1 degree in RF phase. The RF phase stability is influenced by a RF reference line, RF transmission lines, and a RF control system. The RF reference signal is chosen to be a 300-MHz local oscillator (LO) signal, and a rigid copper coaxial line with temperature control was installed for an RF reference distribution. A phase stability of ±0.1 degrees was measured under a temperature change of ±0.1 °C. A digital feedback control system with a field-programmable gate-array (FPGA) module was adopted for a high RF stability. The RF phase was maintained within ±0.1 degrees with a dummy cavity and was within ±0.3 degrees at RFQ operation. In the case of the 20-MeV DTL tanks, one klystron drives 4 tanks, and the input phases of 4 tanks were designed to be in phase. The input phases of 4 tanks were fixed within ±1 degree by adjusting a phase shifter in each waveguide.

  19. Influence of operational conditions on the stability of aerobic granules from the perspective of quorum sensing.

    Science.gov (United States)

    Zhang, Chen; Sun, Supu; Liu, Xiang; Wan, Chunli; Lee, Duu-Jong

    2017-01-25

    Integrated aerobic granules were first cultivated in two sequencing batch reactors (SBRs) (A1 and A2). Then, A1's influent organic loading rate (OLR) was changed from alternating to constant (cycling time was still 6 h), while A2's cycling time varied from 6 to 4 h (influent OLR strategy remained alternating). After 30-day operation since the manipulative alternations, granule breakage happened in two reactors at different operational stages, along with the decrease of granule intensity. Granule diameter in A1 declined from the original 0.84 to 0.32 cm during the whole operation, while granules in A2 dwindled to 0.31 cm on day 22 with similar size to A1. Both the amount of total extracellular polymeric substances (EPSs) and the protein were declining throughout the operation, and the large molecular weight of protein was considered closely related to the stability of aerobic granules. The relative AI-2 level decreased at the same time, and influent OLR strategy might had more evident impact on quorum sensing (QS) ability of sludge compared with starvation period. Combined with microbial results, the decline of total EPS amount in two reactors could be concluded as follows: During the reactor operation, some functional bacteria gradually lost their dominance and were eliminated from the reactors, which finally caused granule disintegration. In summary, the results further confirmed that alternating OLR and proper starvation period were two major factors in effective cultivation and stability of aerobic granules from the perspective of QS.

  20. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models; SUN-RAH: simulador universitario de nucleoelectrica BWR basado en modelos de orden reducido

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2003-07-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  1. Investigation of the Adaptability of Transient Stability Assessment Methods to Real-Time Operation

    DEFF Research Database (Denmark)

    Weckesser, Johannes Tilman Gabriel; Jóhannsson, Hjörtur; Sommer, Stefan

    2012-01-01

    the runtime of each method with respect to the number of inputs. Furthermore, it allows to identify, which method is preferable in case of changes in the power system such as the integration of distributed power resources (DER). A comparison of the performance of the analyzed methods leads to the suggestion......In this paper, an investigation of the adaptability of available transient stability assessment methods to real-time operation and their real-time performance is carried out. Two approaches based on Lyapunov’s method and the equal area criterion are analyzed. The results allow to determine...

  2. Stability Analysis and Stabilization of T-S Fuzzy Delta Operator Systems with Time-Varying Delay via an Input-Output Approach

    Directory of Open Access Journals (Sweden)

    Zhixiong Zhong

    2013-01-01

    Full Text Available The stability analysis and stabilization of Takagi-Sugeno (T-S fuzzy delta operator systems with time-varying delay are investigated via an input-output approach. A model transformation method is employed to approximate the time-varying delay. The original system is transformed into a feedback interconnection form which has a forward subsystem with constant delays and a feedback one with uncertainties. By applying the scaled small gain (SSG theorem to deal with this new system, and based on a Lyapunov Krasovskii functional (LKF in delta operator domain, less conservative stability analysis and stabilization conditions are obtained. Numerical examples are provided to illustrate the advantages of the proposed method.

  3. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  4. Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  5. Inoculum selection is crucial to ensure operational stability in anaerobic digestion.

    Science.gov (United States)

    De Vrieze, Jo; Gildemyn, Sylvia; Vilchez-Vargas, Ramiro; Jáuregui, Ruy; Pieper, Dietmar H; Verstraete, Willy; Boon, Nico

    2015-01-01

    Anaerobic digestion is considered a key technology for the future bio-based economy. The microbial consortium carrying out the anaerobic digestion process is quite complex, and its exact role in terms of "elasticity", i.e., the ability to rapidly adapt to changing conditions, is still unknown. In this study, the role of the initial microbial community in terms of operational stability and stress tolerance was evaluated during a 175-day experiment. Five different inocula from stable industrial anaerobic digesters were fed a mixture of waste activated sludge and glycerol. Increasing ammonium pulses were applied to evaluate stability and stress tolerance. A different response in terms of start-up and ammonium tolerance was observed among the different inocula. Methanosaetaceae were the dominant acetoclastic methanogens, yet, Methanosarcinaceae increased in abundance at elevated ammonium concentrations. A shift from a Firmicutes to a Proteobacteria dominated bacterial community was observed in failing digesters. Methane production was strongly positively correlated with Methanosaetaceae, but also with Bacteria related to Anaerolinaceae, Clostridiales, and Alphaproteobacteria. Volatile fatty acids were strongly positively correlated with Betaproteobacteria and Bacteroidetes, yet ammonium concentration only with Bacteroidetes. Overall, these results indicate the importance of inoculum selection to ensure stable operation and stress tolerance in anaerobic digestion.

  6. Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

    Directory of Open Access Journals (Sweden)

    V. H. Sánchez

    2012-01-01

    Full Text Available The Karlsruhe Institute of Technology (KIT is participating on (Code Applications and Maintenance Program CAMP of the US Nuclear Regulatory Commission (NRC to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test BFBT and plant data recorded during a turbine trip event (TUSA occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.

  7. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    Energy Technology Data Exchange (ETDEWEB)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment.

  8. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  9. A Novel Grid-Wide Transient Stability Assessment and Visualization Method for Increasing Situation Awareness of Control Room Operators

    DEFF Research Database (Denmark)

    Pertl, Michael; Rezkalla, Michel M.N.; Marinelli, Mattia

    2016-01-01

    The aim of the paper is to introduce a grid-wide assessment method to determine the transient stability margin and visualize it effectively to increase the situation awareness of control room operators. Critical area(s) with insufficient transient stability margin have to be identified in order...... to be able to take appropriate preventive actions. The introduced method evaluates the transient stability margin with a time-domain approach by using the voltage angle of several buses across the power system. Information about the severity of a contingency and the location of the most critical buses...... is derived. Moreover, it is shown that the method facilitates the visual examination of transient stability. It provides control room operators with essential information about the state of the system and enables them to take appropriate preventive actions if insufficient transient stability margins...

  10. Oxide evolution on Alloy X-750 in simulated BWR environment

    Science.gov (United States)

    Tuzi, Silvia; Göransson, Kenneth; Rahman, Seikh M. H.; Eriksson, Sten G.; Liu, Fang; Thuvander, Mattias; Stiller, Krystyna

    2016-12-01

    In order to simulate the environment experienced by spacer grids in a boiling water reactor (BWR), specimens of the Ni-based Alloy X-750 were exposed to a water jet in an autoclave at a temperature of 286 °C and a pressure of 80 bar. The oxide microstructure of specimens exposed for 2 h, 24 h, 168 h and 840 h has been investigated mainly using electron microscopy. The specimens suffer mass loss due to dissolution during exposure. At the same time a complex layered oxide develops. After the longest exposure the oxide consists of two outer spinel layers consisting of blocky crystals, one intermediate layer of nickel oxide interspersed with Ti-rich oxide needles, and an inner layer of oxidized base metal. The evolution of the oxide leading up to this structure is discussed and a model is presented.

  11. ON ALGORITHMS CREATION FOR STRAPDOWN STABILIZED GYROCOMPASS OPERATION BASED ON ELECTRICALLY SUSPENDED GYROSCOPE

    Directory of Open Access Journals (Sweden)

    Gennady I. Emelyantsev

    2014-09-01

    Full Text Available The paper presents operation algorithms of ESG-based strapdown stabilized gyrocompass (SSGC located onboard a mobile vehicle. Initial alignment mode and calibration of drift model coefficients onboard a vehicle is aided by reference attitude (heading, pitch and roll angles and position data. The required reference attitude parameters can be generated by joint processing of data from MEMS IMU with gyros and accelerometers and GPS compass. Depending on the vehicle dynamics and required accuracy of generated heading, the system may use IMU based on the fiber-optic gyros and accelerometers instead of MEMS to construct the place vertical. Specific features of SSGC algorithms in heading generation are considered. Calibration and corrected operation modes of the system are described. The SSGC uses two ESGs with orthogonal angular momentum vectors, where one gyro is the reference (unit vector of its angular momentum is aligned with the celestial axis and the other one is virtual (with zero misalignments with respect to the inertial frame, and zero drift model coefficients. Joint processing of SSGC data and external position aiding is realized by extended Kalman filter with full-state feedback control. Simulation modeling results of the system operation algorithms are presented. Simulation modeling has confirmed the system compass effect and determined the time required for calibration of ESG with solid-rotor. The results have been applied at «Concern CSRI “Elektropribor”», JSC

  12. Improvements on the Stability and Operation of a Magnetron H- Ion Source

    Energy Technology Data Exchange (ETDEWEB)

    Sosa, A. [Fermilab; Bollinger, D. S. [Fermilab; Karns, P. R. [Fermilab; Tan, C. Y. [Fermilab

    2017-01-06

    The magnetron H- ion sources developed in the 1970s currently in operation at Fermilab provide beam to the rest of the accelerator complex. A series of modifications to these sources have been tested in a dedicated offline test stand with the aim of improving different operational issues. The solenoid type gas valve was tested as an alternative to the piezoelectric gas valve in order to avoid its temperature dependence. A new cesium oven was designed and tested in order to avoid glass pieces that were present with the previous oven, improve thermal insulation and fine tune its temperature. A current-regulated arc modulator was developed to run the ion source at a constant arc current, providing very stable beam outputs during operations. In order to reduce beam noise, the addition of small amounts of N2 gas was explored, as well as testing different cathode shapes with increasing plasma volume. This paper summarizes the studies and modifications done in the source over the last three years with the aim of improving its stability, reliability and overall performance.

  13. Mine stability evaluation of panel 1 during waste emplacement operations at WIPP

    Energy Technology Data Exchange (ETDEWEB)

    Maleki, H. [Maleki Technologies Inc., Spokane, WA (United States)

    1998-07-01

    The specific objectives of the work were defined by the Environmental Evaluation Group (EEG) as follows: (1) assess the stability of panel 1 during the proposed operation of waste emplacement; (2) estimate the amount of time before room closure would be expected to transfer rock loads to the waste packages. The work consisted of (1) an analysis of geotechnical data and a review of the Department of Energy`s (DOE) plans for waste emplacement in panel 1, (2) an evaluation of ground conditions based on data analysis and observations of changes in ground conditions since the first evaluation in 1993 (USBM 1993), and (3) preparation of a report and presentation of the results to EEG staff. Excluded from this study are radiological safety issues and policies. The study is based on data provided by DOE and Westinghouse Electric Corporation (operator of the site) and conversations with DOE and Westinghouse personnel. MTI cannot independently verify the accuracy of the data within the scope of this study and recommends independent evaluations of data gathering, quality assurance procedures, and structural designs. The operator has the ultimate responsibility for structural designs and has expressed a strong commitment to ensuring worker safety.

  14. BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  15. Operating conditions and stability of spin torque majority gates: Analytical understanding and numerical evidence

    Science.gov (United States)

    Vaysset, Adrien; Manfrini, Mauricio; Nikonov, Dmitri E.; Manipatruni, Sasikanth; Young, Ian A.; Radu, Iuliana P.; Thean, Aaron

    2017-01-01

    The functionality of a cross-shaped Spin Torque Majority Gate (STMG) is primarily limited by the pinning of a domain wall (DW) at the center of the device. Here, an analytical model is built to calculate the conditions for such a pinning and to deduce the operating range. The assumptions of the model and the conclusions are validated by micromagnetic simulations. The total magnetic energy of the DW state is derived. By minimizing this energy with respect to two degrees of freedom, the DW stability condition is obtained. We find that the lateral length of the STMG is the critical dimension: it must be smaller than about five times the DW width. This result is confirmed by micromagnetic simulations with a high accuracy. In process, we solved a more fundamental problem: the macrospin limit of a finite ferromagnet containing one pinning site. We found the correction of the usual DW width expression due to finite length of wires.

  16. Stability of Switched Feedback Time-Varying Dynamic Systems Based on the Properties of the Gap Metric for Operators

    Directory of Open Access Journals (Sweden)

    M. De la Sen

    2012-01-01

    Full Text Available The stabilization of dynamic switched control systems is focused on and based on an operator-based formulation. It is assumed that the controlled object and the controller are described by sequences of closed operator pairs (L,C on a Hilbert space H of the input and output spaces and it is related to the existence of the inverse of the resulting input-output operator being admissible and bounded. The technical mechanism addressed to get the results is the appropriate use of the fact that closed operators being sufficiently close to bounded operators, in terms of the gap metric, are also bounded. That philosophy is followed for the operators describing the input-output relations in switched feedback control systems so as to guarantee the closed-loop stabilization.

  17. Stability of ideal MHD configurations. I. Realizing the generality of the G operator

    Science.gov (United States)

    Keppens, R.; Demaerel, T.

    2016-12-01

    A field theoretical approach, applied to the time-reversible system described by the ideal magnetohydrodynamic (MHD) equations, exposes the full generality of MHD spectral theory. MHD spectral theory, which classified waves and instabilities of static or stationary, usually axisymmetric or translationally symmetric configurations, actually governs the stability of flowing, (self-)gravitating, single fluid descriptions of nonlinear, time-dependent idealized plasmas, and this at any time during their nonlinear evolution. At the core of this theory is a self-adjoint operator G , discovered by Frieman and Rotenberg [Rev. Mod. Phys. 32, 898 (1960)] in its application to stationary (i.e., time-independent) plasma states. This Frieman-Rotenberg operator dictates the acceleration identified by a Lagrangian displacement field ξ , which connects two ideal MHD states in four-dimensional space-time that share initial conditions for density, entropy, and magnetic field. The governing equation reads /d 2 ξ d t 2 = G [ ξ ] , as first noted by Cotsaftis and Newcomb [Nucl. Fusion, Suppl. Part 2, 447 and 451 (1962)]. The time derivatives at left are to be taken in the Lagrangian way, i.e., moving with the flow v. Physically realizable displacements must have finite energy, corresponding to being square integrable in the Hilbert space of displacements equipped with an inner product rule, for which the G operator is self-adjoint. The acceleration in the left-hand side features the Doppler-Coriolis operator v . ∇ , which is known to become an antisymmetric operator when restricting attention to stationary equilibria. Here, we present all derivations needed to get to these insights and connect results throughout the literature. A first illustration elucidates what can happen when self-gravity is incorporated and presents aspects that have been overlooked even in simple uniform media. Ideal MHD flows, as well as Euler flows, have essentially 6 + 1 wave types, where the 6 wave modes

  18. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 {sup o}C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV

  19. Estimation of the coolant flow through a natural circulation BWR fuel channel applying and equivalent electrical model

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Campus Morelos en IMTA, Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_ig@yahoo.com.m [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    This work presents the design and implementation of an advanced controller for a reduced order model of a BWR reactor core cooled by natural circulating water, which allows real time estimates of coolant flows through fuel assemblies about standard neutron flux strings. Nuclear power plants with boiling water reactors control individual fuel assembly coolant flows by forced circulation using external or internal water pumps and different core support plate orifices. These two elements reduce flow dependency on local channel pressure drops. In BWR reactors using only natural circulation coolant flows, these two elements are not available and therefore individual channel coolant flows are highly dependent in local conditions, such as power distributions and local pressure drops. Therefore it is expected that grater uncertainties in these variables be used during safety, fuel management and other analysis, which in turns may lead to increased operation penalties, such as tighter operating limits. The objective of this work is to asses by computer simulations means to reduce uncertainties in the measurement of fuel assembly coolant flows and eventually the associated penalties. During coolant phase transitions, pressure drops and local power may alter local natural circulation through fuel assemblies and flow estimates can be helped or not by control rod moves. This work presents the construction of an optimal controller for a core flow estimator based on a reduced order model of the coolant going though the reactor vessel components and nuclear core. This model is to be driven by plant signals from standard BWR instrumentation in order to estimate the coolant flows in selected fuel assemblies about a LPRM string. For this purpose an equivalent electrical model has been mathematically developed and numerically tested. The power-flow maps of typical BRW are used as steady state references for this equivalent model. Once these were fully reproduced for steady state

  20. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  1. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR; BUTREN-RC un sistema hibrido para la optimizacion de recargas de combustible nuclear en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Castillo M, J.A. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  2. USING THE SULFUR POLYMER STABILIZATION SOLIDIFICATION PROCESS TO TREAT RESIDUAL MERCURY WASTES FROM GOLD MINING OPERATIONS.

    Energy Technology Data Exchange (ETDEWEB)

    BOWERMAN,B.ADAMS,J.KALB,P.WAN,R.Y.LEVIER,M.

    2003-02-24

    Large quantities of mercury are generated as a by-product during the processing of gold ore following mining operations. Newmont Mining Corporation (NMC), which operates some of the world's largest gold mines, sought a method to permanently ''retire'' its mercury by-products, thereby avoiding potential environmental liability. Sulfur Polymer Stabilization-Solidification (SPSS) is an innovative technology developed at Brookhaven National Laboratory (BNL) for treatment of mercury and mercury contaminated materials, such as soil, sludge and debris. BNL conducted a treatability study to determine the potential applicability of SPSS for treatment of Newmont mercury, and the treated product passed the U.S. Environmental Protection Agency (EPA) test for toxicity. The SPSS process has been shown to be effective on radioactive and nonradioactive mercury and mercury-contaminated materials with a pilot-scale batch system capable of producing 0.03 m{sup 3} (1 ft{sup 3}) per batch. Engineering scale-up issues are discussed and material property tests addressing these issues are described.

  3. Impact of MSWI Bottom Ash Codisposed with MSW on Landfill Stabilization with Different Operational Modes

    Directory of Open Access Journals (Sweden)

    Wen-Bing Li

    2014-01-01

    Full Text Available The aim of the study was to investigate the impact of municipal solid waste incinerator (MSWI bottom ash (BA codisposed with municipal solid waste (MSW on landfill stabilization according to the leachate quality in terms of organic matter and nitrogen contents. Six simulated landfills, that is, three conventional and three recirculated, were employed with different ratios of MSWI BA to MSW. The results depicted that, after 275-day operation, the ratio of MSWI BA to fresh refuse of 1 : 10 (V : V in the landfill was still not enough to provide sufficient acid-neutralizing capacity for a high organic matter composition of MSW over 45.5% (w/w, while the ratio of MSWI BA to fresh refuse of 1 : 5 (V : V could act on it. Among the six experimental landfills, leachate quality only was improved in the landfill operated with the BA addition (the ratio of MSWI BA to fresh refuse of 1 : 5 (V : V and leachate recirculation.

  4. Performance, stability and operation voltage optimization of screen-printed aqueous supercapacitors

    Science.gov (United States)

    Lehtimäki, Suvi; Railanmaa, Anna; Keskinen, Jari; Kujala, Manu; Tuukkanen, Sampo; Lupo, Donald

    2017-01-01

    Harvesting micropower energy from the ambient environment requires an intermediate energy storage, for which printed aqueous supercapacitors are well suited due to their low cost and environmental friendliness. In this work, a systematic study of a large set of devices is used to investigate the effect of process variability and operating voltage on the performance and stability of screen printed aqueous supercapacitors. The current collectors and active layers are printed with graphite and activated carbon inks, respectively, and aqueous NaCl used as the electrolyte. The devices are characterized through galvanostatic discharge measurements for quantitative determination of capacitance and equivalent series resistance (ESR), as well as impedance spectroscopy for a detailed study of the factors contributing to ESR. The capacitances are 200–360 mF and the ESRs 7.9–12.7 Ω, depending on the layer thicknesses. The ESR is found to be dominated by the resistance of the graphite current collectors and is compatible with applications in low-power distributed electronics. The effects of different operating voltages on the capacitance, leakage and aging rate of the supercapacitors are tested, and 1.0 V found to be the optimal choice for using the devices in energy harvesting applications. PMID:28382962

  5. On Stability of Open-Loop Operation without Rotor Information for Brushless DC Motors

    Directory of Open Access Journals (Sweden)

    Zhong Wu

    2014-01-01

    Full Text Available Open-loop operation mode is often used to control the Brushless DC Motors (BLDCMs without rotor position sensors when the back electromotive force (EMF is too weak due to the very low rotor velocity. The rotor position information is not necessary in this mode and the stator windings are supplied with voltages under a certain ratio of the amplitude to the frequency. However, the rotor synchronization will be destroyed once if the commutation instant is inappropriate. In order to improve the reliability of the open-loop operation mode, a dynamic equation is established to represent the synchronization error between the rotor and the stator. Thereafter, the stability of the open-loop control mode is analyzed by using Lyapunov indirect method. Theoretical analysis indicates that the open-loop control mode is asymptotically stable only when the commutation instant of the stator current lags behind the ideal one suitably. Finally, theoretical analysis is verified through the experimental results of a certain BLDCM.

  6. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  7. Plant analyzer for high-speed interactive simulation of BWR power plant transients

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H.S.; Lekach, S.V.; Mallen, A.N.; Wulff, W.; Cerbone, R.J.

    1984-04-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times faster than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feedwater train. Point kinetics incorporate reactivity feedback due to void fraction, fuel temperature, coolant temperature, and boron concentration. Control systems and trip logic are simulated for the nuclear steam supply system. The AD10 of Applied Dynamics International is the special-purpose peripheral processor. It is specifically designed for high-speed digital system simulation, accommodates hardware (instrumentation) in the input/output loop, and operates interactively on-line, like an analog computer. Results are shown to demonstrate computing capacity, accuracy, and speed. Simulation speeds have been achieved which are orders of magnitude faster than those of a CDC-7600 mainframe computer or ten times faster than real-time speed.

  8. Generic BWR-4 degraded core in-vessel study. Status report

    Energy Technology Data Exchange (ETDEWEB)

    1984-11-01

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination.

  9. Diffusion bonded matrix of HGMF applied for BWR condensate water purification

    Energy Technology Data Exchange (ETDEWEB)

    Soda, Fumitaka; Yukawa, Takao; Ito, Kazuyuki

    1984-03-01

    A high Gradient Magnetic Filter (HGMF) applied to the purification of power plant primary water has recently attracted much attention. In the application of HGMF to the water treatment of power plants, especially nuclear power plants, reliabillties of matrix (filtering medium) as well as removal performance for cruds (insoluble corrosion products) are considered to be important factors. To satisfy these factors, a new filtering medium named Diffision Bonded Matrix (DBM) has been developed and the test results are reported. Filtering efficiency and mechanical stiffness of DBM were examined using HGMF pilot test units consisting of 160 mm diameters x 240 mm length filter. The filtering velocity and the magnetic flux density used in this test were 800 m/h 5 kG, respectively. The filtering efficiencies and of 85-100% were obtained for artificial cruds for DBM. The DBM indicated slightly better filtering efficiency than for conventional wool matrix under the same filtering and matrix conditions. The DBM kept its original mechanical properties and very few pieces of fibers were broken off while the conventional wool matrix lost its volume elasticities and the considerable amount of fibers was broken off during the test operation. The results described here demonstrated the applicability of DBM for treatment of BWR primary water by High Gradient Magnetic Filter.

  10. In-plant material test experience under hydrogen water chemistry at a Japanese BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Masami; Koshiishi, Masato; Kato, Takahiko [Hitachi Ltd., Ibaraki (Japan). Hitachi Works; Abe, Ayumi; Sekiguchi, Masahiko; Takiguchi, Hideki

    1999-07-01

    Hydrogen injection technology has been applied to Japanese domestic aged BWR plants since 1994 to mitigate corrosive environment regarding Intergranular Stress Corrosion Cracking (IGSCC) of Reactor Internals (RINs). The Tsuruga Unit-1 plant has also been operated with this technology since 1997, considering suppression of radiation increase in the main steam piping system besides mitigation of corrosive environment in the reactor; the hydrogen injection rate in the feed water was about 0.5 ppm. In order to confirm the effects of the hydrogen injection on suppression of SCC susceptibility of the RIN materials, several in-plant material tests have been conducted using the reactor water clean up system (RWCU). Cyclic-Slow Strain Rate Tensile (C-SSRT) test, Slow Strain Rate Tensile (SSRT) test and Compact Tension (CT) test were performed in the test facilities which were installed at the sampling line from the RWCU. Evaluation of SCC life by means of the C-SSRT test was the first application as an accelerated SCC test for in-plant material tests. It was confirmed that the hydrogen injection in the feed water has a good mitigation effects on IGSCC performance of the RIN materials. Results will be discussed from a viewpoint of the test condition such as total oxidant, ECP, conductivity and loading/unloading. (author)

  11. Analysis of the behavior of irradiated BWR fuel rod in storage dry conditions; Analisis del comportamiento de una barra combustible irradiada BWR en condiciones de almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Montes, D.; Ruiz-Hervias, J.; Munoz-Reja, C.

    2014-07-01

    In order to complete previous studies of creep on PWR sheath material, developed a joint experimental program by CSN, ENRESA and ENUSA about BWR (Zircaloy-2) sheath material. This program consisted in creep tests and then on the material under creep, compression testing diametral obtaining the permissible displacement of the sheath to break. (Author)

  12. Study of instabilities in phase by using the tool {sup D}ynamics{sup :} analysis of the evolution space temporary of the waves of density in channels of reactors BWR; Estudio de las Inestabilidades en Fase Mediante la Herramienta Dinamics: analisis de la Evolucion Espacio Temporal de las Ondas de Densidad en Canales de Reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Escriva, R.; Merino, R.; Melara, J.

    2013-07-01

    This paper presents the basics of Dynamics V2 to code It allows calculations of stability for oscillations in phase in BWR reactors in the time domain. The equations of the model are exposed and is the integration of the equations. The model can be used in a large number of nodes thrust for the calculations to an acceptable computational cost, it has simplified dynamics of recirculation loop and the code has been incorporated the Oscillation in phase boundary conditions. The code incorporates the equations of boiling sub-cooled which allows to make more realistic calculations as well as subroutines to calculate the subroutines-based properties of the MATPRO and ASME.

  13. Description and assessment of RAMONA-3B Mod. 0 Cycle 4: a computer code with three-dimensional neutron kinetics for BWR system transients

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W; Cheng, H S; Diamond, D J; Khatib-Rahbar, M

    1984-01-01

    This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future.

  14. Potential effects of desalinated water quality on the operation stability of wastewater treatment plants.

    Science.gov (United States)

    Lew, Beni; Cochva, Malka; Lahav, Ori

    2009-03-15

    Desalinated water is expected to become the major source of drinking water in many places in the near future, and thus the major source of wastewater to arrive at wastewater treatment plants. The paper examines the effect of the alkalinity value with which the water is released from the desalination plant on the alkalinity value that would develop within the wastewater treatment process under various nitrification-denitrification operational scenarios. The main hypothesis was that the difference in the alkalinity value between tap water and domestic wastewater is almost exclusively a result of the hydrolysis of urea (NH(2)CONH(2), excreted in the human urine) to ammonia (NH(3)), regardless of the question what fraction of NH(3(aq)) is transformed to NH(4)(+). Results from a field study show that the ratio between the alkalinity added to tap water when raw wastewater is formed (in meq/l units) and the TAN (total ammonia nitrogen, mole/l) concentration in the raw wastewater is almost 1:1 in purely domestic sewage and close to 1:1 in domestic wastewater streams mixed with light industry wastewaters. Having established the relationship between TAN and total alkalinity in raw wastewater the paper examines three theoretical nitrification-denitrification treatment scenarios in the wastewater treatment plant (WWTP). The conclusion is that if low-alkalinity desalinated water constitutes the major water source arriving at the WWTP, external alkalinity will have to be added in order to avoid pH drop and maintain process stability. The results lead to the conclusion that supplying desalinated water with a high alkalinity value (e.g. > or =100 mg/l as CaCO(3)) would likely prevent the need to add costly basic chemicals in the WWTP, while, in addition, it would improve the chemical and biological stability of the drinking water in the distribution system.

  15. An integrated approach to monitoring the calibration stability of operational dual-polarization radars

    Science.gov (United States)

    Vaccarono, Mattia; Bechini, Renzo; Chandrasekar, Chandra V.; Cremonini, Roberto; Cassardo, Claudio

    2016-11-01

    The stability of weather radar calibration is a mandatory aspect for quantitative applications, such as rainfall estimation, short-term weather prediction and initialization of numerical atmospheric and hydrological models. Over the years, calibration monitoring techniques based on external sources have been developed, specifically calibration using the Sun and calibration based on ground clutter returns. In this paper, these two techniques are integrated and complemented with a self-consistency procedure and an intercalibration technique. The aim of the integrated approach is to implement a robust method for online monitoring, able to detect significant changes in the radar calibration. The physical consistency of polarimetric radar observables is exploited using the self-consistency approach, based on the expected correspondence between dual-polarization power and phase measurements in rain. This technique allows a reference absolute value to be provided for the radar calibration, from which eventual deviations may be detected using the other procedures. In particular, the ground clutter calibration is implemented on both polarization channels (horizontal and vertical) for each radar scan, allowing the polarimetric variables to be monitored and hardware failures to promptly be recognized. The Sun calibration allows monitoring the calibration and sensitivity of the radar receiver, in addition to the antenna pointing accuracy. It is applied using observations collected during the standard operational scans but requires long integration times (several days) in order to accumulate a sufficient amount of useful data. Finally, an intercalibration technique is developed and performed to compare colocated measurements collected in rain by two radars in overlapping regions. The integrated approach is performed on the C-band weather radar network in northwestern Italy, during July-October 2014. The set of methods considered appears suitable to establish an online tool to

  16. Charging of capacitors with double switch. The principle of operation of auto-zero and chopper-stabilized DC amplifiers

    CERN Document Server

    Yordanov, Vasil G; Manolev, Stojan G; Mishonov, Todor M

    2015-01-01

    The principle of operation of auto-zero and chopper-stabilized DC amplifiers, which is realized in many contemporary operational amplifiers is illustrated by a simple experimental setup given at the Open Experimental Physics Olympiad 2014 - "The Day of the Capacitor", held in Sofia and Gevgelija. The Olympiad was organized by the Sofia Branch of the Union of Physicists in Bulgaria and the Regional Society of Physicists of Strumica, Macedonia. In addition to the solution of the secondary school task in the paper is given a detailed engineering description of the patent by Edwin Goldberg and Jules Lehmann, Stabilized direct current amplifier, U.S. Patent 2,684,999 (1949).

  17. Stability Transfer between Two Clock Lasers Operating at Different Wavelengths for Absolute Frequency Measurement of Clock Transition in 87Sr

    CERN Document Server

    Yamaguchi, A; Nagano, S; Li, Y; Ishijima, H; Hachisu, H; Kumagai, M; Ido, T; 10.1143/APEX.5.022701

    2012-01-01

    We demonstrated transferring the stability of one highly stable clock laser operating at 729 nm to another less stable laser operating at 698 nm. The two different wavelengths were bridged using an optical frequency comb. The improved stability of the clock laser at 698 nm enabled us to evaluate the systematic frequency shifts of the Sr optical lattice clock with shorter averaging time. We determined the absolute frequency of the clock transition 1S0 - 3P0 in 87Sr to be 429 228 004 229 873.9 (1.4) Hz referenced to the SI second on the geoid via International Atomic Time (TAI).

  18. Stabilization

    Directory of Open Access Journals (Sweden)

    Muhammad H. Al-Malack

    2016-07-01

    Full Text Available Fuel oil flyash (FFA produced in power and water desalination plants firing crude oils in the Kingdom of Saudi Arabia is being disposed in landfills, which increases the burden on the environment, therefore, FFA utilization must be encouraged. In the current research, the effect of adding FFA on the engineering properties of two indigenous soils, namely sand and marl, was investigated. FFA was added at concentrations of 5%, 10% and 15% to both soils with and without the addition of Portland cement. Mixtures of the stabilized soils were thoroughly evaluated using compaction, California Bearing Ratio (CBR, unconfined compressive strength (USC and durability tests. Results of these tests indicated that stabilized sand mixtures could not attain the ACI strength requirements. However, marl was found to satisfy the ACI strength requirement when only 5% of FFA was added together with 5% of cement. When the FFA was increased to 10% and 15%, the mixture’s strength was found to decrease to values below the ACI requirements. Results of the Toxicity Characteristics Leaching Procedure (TCLP, which was performed on samples that passed the ACI requirements, indicated that FFA must be cautiously used in soil stabilization.

  19. BWR: Development and Validation of KERENA reactor; Les REB: Developpement et validation du reacteur KERENA

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, F.; Fuchs, M. [E.ON Kernkraft GmbH (Germany); Erve, M.; Pasler, D. [AREVA (Germany)

    2010-07-01

    KERENA is an advanced boiling water reactor, combining AREVA's and E.ON's expertise. A project was launched to customize the final basic design for this advanced nuclear power plant having a net power output of about 1, 250 MW, a net efficiency of about 37% and a design service life of 60 years. The development takes into account the technical and accumulated operating experience of the project partners. The plant safety concept is based on an optimized combination of a reduced number of proven active safety systems and passive safety systems, utilizing basic laws of physics, such as gravity, enabling them to function without electrical power supplies or activation by powered instrumentation and control systems. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. All passive safety systems are validated in an experimental test program at AREVA, using 1:1 scale test facilities (INKA test facility Karlstein). The KERENA boiling water reactor is compliant with international nuclear codes and standards, and is also designed to withstand the effects of an aircraft crash involving a military aircraft or a large passenger airline. The safety level of the KERENA reactor has been able to be significantly increased compared to existing BWR plants. The advantages of the new safety concept are: -) Reduced susceptibility of safety systems to failures; -) Larger safety margins; -) Good plant behavior in the event of accidents due to the fact that conditions change at a slower rate; -) Grace periods of several days after an accident before operator intervention is required; -) Significantly reduced impact of operator error on reactor safety; -) No need for large-scale emergency response actions such as temporary evacuation or relocation of the neighboring population following a core melt accident. (A.C.)

  20. Environmental mitigation for SCC initiation of BWR core internals by hydrogen injection during start-up

    Energy Technology Data Exchange (ETDEWEB)

    Dozaki, K.; Abe, A.; Nagata, N.; Takiguchi, H. [The Japan Atomic Power Co. (Japan)

    2004-07-01

    Hydrogen injection into the reactor water has been applied to many BWR power stations. Since hydrogen injected accelerates recombination of oxidant generated by water radiolysis, oxidant concentration, such as dissolved oxygen concentration in reactor water can be reduced. As the result of the reduction of oxidant concentration, Electrochemical Corrosion Potential (ECP) at the surface of structural material can be lowered. Lowered ECP moderates Stress Corrosion Cracking (SCC) sensitivity of structural materials, such as stainless steels. As usual, hydrogen injection system begins to work after the plant start-up is finished, when the condition of normal operation is established. Accordingly, Hydrogen Water Chemistry (HWC) does not cover all the period of plant operation. As far as SCC crack growth is considered, loss of HWC during plant start-up does not result in significant crack growth, because of duration of plant start-up is much shorter than that of plant normal operation, when HWC condition is being satisfied. However, the reactor water environment and load conditions during a plant start-up may contribute to the initiation of SCC. It is estimated that the core internals are subjected to the strain rate that may cause susceptibility to SCC initiation during start-up. Dissolved oxygen (DO) and hydrogen peroxide (H{sub 2}O{sub 2}) has a peak, and ECP is in high levels during start-up. Therefore it is beneficial to perform hydrogen injection during start-up as well in order to suppress SCC initiation. We call it HWC During Start-up (HDS) here. (orig.)

  1. Erythropoietin improves operant conditioning and stability of cognitive performance in mice

    Directory of Open Access Journals (Sweden)

    Ehrenreich Hannelore

    2009-07-01

    Full Text Available Abstract Background Executive functions, learning and attention are imperative facets of cognitive performance, affected in many neuropsychiatric disorders. Recently, we have shown that recombinant human erythropoietin improves cognitive functions in patients with chronic schizophrenia, and that it leads in healthy mice to enhanced hippocampal long-term potentiation, an electrophysiological correlate of learning and memory. To create an experimental basis for further mechanistic insight into erythropoietin-modulated cognitive processes, we employed the Five Choice Serial Reaction Time Task. This procedure allows the study of the effects of erythropoietin on discrete processes of learning and attention in a sequential fashion. Results Male mice were treated for 3 weeks with erythropoietin (5,000 IU/kg versus placebo intraperitoneally every other day, beginning at postnatal day 28. After termination of treatment, mice were started on the Five Choice Serial Reaction Time Task, with daily training and testing extending to about 3 months. Overall, a significantly higher proportion of erythropoietin-treated mice finished the task, that is, reached the criteria of adequately reacting to a 1.0 sec flash light out of five arbitrarily appearing choices. During acquisition of this capability, that is, over almost all sequential training phases, learning readouts (magazine training, operant and discriminant learning, stability of performance were superior in erythropoietin-treated versus control mice. Conclusion Early erythropoietin treatment leads to lasting improvement of cognitive performance in healthy mice. This finding should be exploited in novel treatment strategies for brain diseases.

  2. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  3. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  4. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  5. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  6. Optimization of operation conditions for preventing sludge bulking and enhancing the stability of aerobic granular sludge in sequencing batch reactors.

    Science.gov (United States)

    Zhou, Jun; Wang, Hongyu; Yang, Kai; Ma, Fang; Lv, Bin

    2014-01-01

    Sludge bulking caused by loss of stability is a major problem in aerobic granular sludge systems. This study investigated the feasibility of preventing sludge bulking and enhancing the stability of aerobic granular sludge in a sequencing batch reactor by optimizing operation conditions. Five operation parameters have been studied with the aim to understand their impact on sludge bulking. Increasing dissolved oxygen (DO) by raising aeration rates contributed to granule stability due to the competition advantage of non-filamentous bacteria and permeation of oxygen at high DO concentration. The ratio of polysaccharides to proteins was observed to increase as the hydraulic shear force increased. When provided with high/low organic loading rate (OLR) alternately, large and fluffy granules disintegrated, while denser round-shape granules formed. An increase of biomass concentration followed a decrease at the beginning, and stability of granules was improved. This indicated that aerobic granular sludge had the resistance of OLR. Synthetic wastewater combined highly and slowly biodegradable substrates, creating a high gradient, which inhibited the growth of filamentous bacteria and prevented granular sludge bulking. A lower chemical oxygen demand/N favored the hydrophobicity of granular sludge, which promoted with granule stability because of the lower diffusion rate of ammonia. The influence of temperature indicated a relatively low temperature was more suitable.

  7. Fast determination of operational stability of the soluble acetylacetone-cleaving enzyme Dke1 in an enzyme membrane reactor.

    Science.gov (United States)

    Hofer, Hannes; Steiner, Walter

    2005-11-01

    The main aim of this study was the determination of the operational stability of soluble Dke1 (EC 1.13.11.50) in an enzyme membrane reactor. In order to calculate the half-life of soluble Dke1, the K (M) of oxygen must be known. The determination of this constant was done using progress curve analysis (K (M) = 260 micromol l(-1)). In a next step, the reactor system was studied by building a mathematical model for calculation of the reactor system, using Berkeley Madonna ver. 8.0.1 software. After that, the determination of the half-life of Dke1 under operational conditions at different temperatures (5, 10, 15, 25, 30, 35 degrees C) was performed. The quantitative criterion for stability was the value of the first-order rate constant of monomolecular inactivation. The experiments showed that soluble Dke1 is poorly stable. The half-life ranged from 308 min at 5 degrees C to 9 min at 35 degrees C. This method for determining the half-life is quite applicable for enzymes which are poorly stable. In addition, both the storage stability and the operational stability can be determined.

  8. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  9. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  10. Stabilizing control system preventing loss of synchronism from extension and its actual operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Matsuzawa, K.; Yanagihashi, K.; Tsukita, J. [Tokyo Electric Power Co. (Japan); Sato, M.; Nakamura, T.; Takeuchi, A.

    1995-08-01

    This paper presents a stabilizing control system that was implemented to the power system of Tokyo Electric Power Company and worked successfully, shedding the minimum number of generators and preventing the extension of loss of synchronism to the other generators when a severe fault occurred on May 22, 1992. The stabilizing control system measures voltage and current at the rate of 600 Hz, predicts the change of phase angles of generators and decides the minimum number of generators to be shed for stabilization when instability is predicted and sheds them.

  11. Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  12. Overview of New Tools to Perform Safety Analysis: BWR Station Black Out Test Case

    Energy Technology Data Exchange (ETDEWEB)

    D. Mandelli; C. Smith; T. Riley; J. Nielsen; J. Schroeder; C. Rabiti; A. Alfonsi; Cogliati; R. Kinoshita; V. Pasucci; B. Wang; D. Maljovec

    2014-06-01

    Dynamic Probabilistic Risk Assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). While system simulator codes accurately model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by: 1) sampling values of a set of parameters from the uncertainty space of interest (using the simulation controller codes), and 2) simulating the system behavior for that specific set of parameter values (using the system simulator codes). For complex systems, one of the major challenges in using DPRA methodologies is to analyze the large amount of information (i.e., large number of scenarios ) generated, where clustering techniques are typically employed to allow users to better organize and interpret the data. In this paper, we focus on the analysis of a nuclear simulation dataset that is part of the Risk Informed Safety Margin Characterization (RISMC) Boiling Water Reactor (BWR) station blackout (SBO) case study. We apply a software tool that provides the domain experts with an interactive analysis and visualization environment for understanding the structures of such high-dimensional nuclear simulation datasets. Our tool encodes traditional and topology-based clustering techniques, where the latter partitions the data points into clusters based on their uniform gradient flow behavior. We demonstrate through our case study that both types of clustering techniques complement each other in bringing enhanced structural understanding of the data.

  13. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  14. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

    2016-05-15

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  15. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  16. Stability of systems with the Duhem hysteresis operator : The dissipativity approach

    NARCIS (Netherlands)

    Jayawardhana, Bayu; Ouyang, Ruiyue; Andrieu, Vincent

    2012-01-01

    In this paper, we discuss the dissipativity property of the counterclockwise Duhem operator. Sufficient conditions on the functions which define the Duhem operator are given such that the Duhem operator has counterclockwise input-output dynamics. In particular, an explicit construction of the storag

  17. Wall-shaped electrodes for reducing the operation voltage of polymer-stabilized blue phase liquid crystal displays

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Miyoung; Kim, Min Su; Kang, Byeong Gyun; Kim, Mi-Kyung; Yoon, Sukin; Lee, Seung Hee [Polymer BIN Fusion Research Center, Department of Polymer Nano-Science and Technology, Chonbuk National University, Chonju, Chonbuk 561-756 (Korea, Republic of); Ge Zhibing; Rao Linghui; Gauza, Sebastian; Wu, Shin-Tson, E-mail: lsh1@chonbuk.ac.k, E-mail: swu@creol.ucf.ed [College of Optics and Photonics, University of Central Florida, Orlando, FL 32816 (United States)

    2009-12-07

    Polymer-stabilized blue phase liquid crystal displays based on the Kerr effect are emerging due to their submillisecond response time, wide view and simple fabrication process. However, the conventional in-plane switching device exhibits a relatively high operating voltage because the electric fields are restricted in the vicinity of the electrode surface. To overcome this technical barrier, we propose a partitioned wall-shaped electrode configuration so that the induced birefringence is uniform between electrodes throughout the entire cell gap. Consequently, the operating voltage is reduced by {approx} 2.8x with two transistors. The responsible physical mechanisms are explained.

  18. The Evolving Contingency Contracting Market: Private Sector Self regulation and United States Government Monitoring of Procurement of Stability Operations Services

    Science.gov (United States)

    2016-01-01

    practice with preaching. Daniel A. Pinnell COL, U.S. Army Director Peacekeeping and Stability Operations Institute iv v ABOUT THE AUTHOR WHITNEY...International Code of Conduct for Private Security Providers, complemented by case studies of private firm initiatives and successes in combating human...activities. Case Study I: The International Code of Conduct for Private Security Providers In 2008, the International Committee of the Red Cross, the

  19. An integer order approximation method based on stability boundary locus for fractional order derivative/integrator operators.

    Science.gov (United States)

    Deniz, Furkan Nur; Alagoz, Baris Baykant; Tan, Nusret; Atherton, Derek P

    2016-05-01

    This paper introduces an integer order approximation method for numerical implementation of fractional order derivative/integrator operators in control systems. The proposed method is based on fitting the stability boundary locus (SBL) of fractional order derivative/integrator operators and SBL of integer order transfer functions. SBL defines a boundary in the parametric design plane of controller, which separates stable and unstable regions of a feedback control system and SBL analysis is mainly employed to graphically indicate the choice of controller parameters which result in stable operation of the feedback systems. This study reveals that the SBL curves of fractional order operators can be matched with integer order models in a limited frequency range. SBL fitting method provides straightforward solutions to obtain an integer order model approximation of fractional order operators and systems according to matching points from SBL of fractional order systems in desired frequency ranges. Thus, the proposed method can effectively deal with stability preservation problems of approximate models. Illustrative examples are given to show performance of the proposed method and results are compared with the well-known approximation methods developed for fractional order systems. The integer-order approximate modeling of fractional order PID controllers is also illustrated for control applications.

  20. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  1. Effect of operation parameters on the flux stabilization of gravity-driven membrane (GDM) filtration system for decentralized water supply.

    Science.gov (United States)

    Tang, Xiaobin; Ding, An; Qu, Fangshu; Jia, Ruibao; Chang, Haiqing; Cheng, Xiaoxiang; Liu, Bin; Li, Guibai; Liang, Heng

    2016-08-01

    A pilot-scale gravity-driven membrane (GDM) filtration system under low gravitational pressure without any pre-treatment, backwash, flushing, or chemical cleaning was carried out to investigate the effect of operation parameters (including operation pressure, aeration mode, and intermittent filtration) on the effluent quality and permeability development. The results revealed that GDM system exhibited an efficient performance for the removal of suspended substances and organic compounds. The stabilization of flux occurred and the average values of stable flux were 6.6, 8.1, and 8.6 Lm(-2) h(-1) for pressures of 65, 120, and 200 mbar, respectively. In contrast, flux stabilization was not observed under continuous and intermittent aeration conditions. However, aeration (especially continuous aeration) was effective to improve flux and alleviate membrane fouling during 1-month operation. Moreover, intermittent filtration would influence the stabilization of permeate flux, resulting in a higher stable flux (ranging from 6 to 13 Lm(-2) h(-1)). The stable flux significantly improved with the increase of intermittent period. Additionally, GDM systems exhibited an efficient recovery of flux after simple physical cleaning and the analyses of resistance reversibility demonstrated that most of the total resistance was hydraulic reversible resistance (50-75 %). Therefore, it is expected that the results of this study can develop strategies to increase membrane permeability and reduce energy consumption in GDM systems for decentralized water supply.

  2. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning; Perfiles axiales de quemado y fraccion de huecos para calculos de criticidad con credito al quemado para combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-07-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  3. Strain-induced corrosion cracking in ferritic components of BWR primary circuits; Risskorrosion in druckfuehrenden ferritischen Komponenten des Primaerkreislaufes von Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S.; Ineichen, U.; Tschanz, U.; Gerodetti, B

    2003-04-01

    The present final report of the RIKORR project is a summary of a literature survey and of the experimental work performed by PSI on the environmentally-assisted cracking (EAC) and dynamic strain ageing (DSA) susceptibility of low-alloy steels (LAS) in high-temperature (HT) water. Within this project, the EAC crack growth behaviour of different low-alloy RPV steels, weld filler and weld heat-affected zone materials has been investigated under simulated transient and steady-state BWR/NWC power operation conditions. The strain-induced corrosion cracking (SICC) / low-frequency corrosion fatigue (CF) and stress corrosion cracking (SCC) crack growth behaviour of different low-alloy RPV steels under simulated transient and stationary BWR/NWC conditions was characterized by slow rising load / low-frequency corrosion fatigue and constant load / periodical partial unloading / ripple load tests with pre-cracked fracture mechanics specimens in oxygenated HT water at temperatures of either 288, 250, 200 or 150 {sup o}C. Modern high-temperature water loops, on-line crack growth monitoring and fractographic analysis by scanning electron microscopy (SEM) were used to quantify the cracking response. (author)

  4. Endurance and stability of some surface meteorological sensors under land- and ship-based operating environments

    Digital Repository Service at National Institute of Oceanography (India)

    Mehra, P.; Desai, R.G.P.; Joseph, A.; VijayKumar, K.; Dabholkar, N.; Prabhudesai, S.; Nagvekar, S.; Agarvadekar, Y.

    ), the ship-based AWS system suffered differing kinds of oscillatory motions (pitch, roll, yaw) under varying sea state conditions. In the present study, performance of the sensors is assessed in terms of two criteria namely; (1) endurance, and (2) stability...

  5. Intra-operative evaluation of cementless hip implant stability: a prototype device based on vibration analysis.

    Science.gov (United States)

    Lannocca, Maurizio; Varini, Elena; Cappello, Angelo; Cristofolini, Luca; Bialoblocka, Ewa

    2007-10-01

    Cementless implants are mechanically stabilized during surgery by a press-fitting procedure. Good initial stability is crucial to avoid stem loosening and bone cracking, therefore, the surgeon must achieve optimal press-fitting. A possible approach to solve this problem and assist the surgeon in achieving the optimal compromise, involves the use of vibration analysis. The present study aimed to design and test a prototype device able to evaluate the primary mechanical stability of a cementless prosthesis, based on vibration analysis. In particular, the goal was to discriminate between stable and quasi-stable implants; thus the stem-bone system was assumed to be linear in both cases. For that reason, it was decided to study the frequency responses of the system, instead of the harmonic distortion. The prototype developed consists of a piezoelectric exciter connected to the stem and an accelerometer attached to the femur. Preliminary tests were performed on four composite femurs implanted with a conventional stem. The results showed that the input signal was repeatable and the output could be recorded accurately. The most sensitive parameter to stability was the shift in resonance frequency of the stem-bone system, which was highly correlated with residual micromotion on all four specimens.

  6. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network; Prediccion del factor local de potencia en celdas de combustible BWR mediante una red neuronal multicapas

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 La Marquesa, Ocoyoacac, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2007-07-01

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U{sup 235}, some of these bars also contain a concentration of Gd{sub 2}O{sub 3} and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  7. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  8. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  9. Improving the Operational Stability of PBDTTTz-4 Polymer Solar Cells Modules by Electrode Modification

    DEFF Research Database (Denmark)

    Roth, Bérenger; Benatto, Gisele Alves dos Reis; Corazza, Michael;

    2016-01-01

    PBDTTTz-4 is employed in the ambient manufacturing of fully Roll-to-Roll organic solar cell modules. Modules are manufactured using a novel silver nanowire electrode or a previously reported carbon electrode. The average PCE of carbon modules (3.07%) and AgNW modules (1.46%) shows that PBDTTTz-4...... is a good candidate for upscaling. Stability measurements following the ISOS standards are used to compare the lifetime of the different modules. In all tests but one, the carbon modules are less stable. The higher stability of AgNW is attributed to the removal of the PEDOT:PSS in the front electrode....... Finally during indoor light tests, a new degradation phenomenon is observed where bubbles are formed inside the modules contrary to previous reports of bubble formation by thermal expansion of trapped gas inside the barrier....

  10. Ulam stability for fractional differential equations in the sense of Caputo operator

    Directory of Open Access Journals (Sweden)

    Rabha W. Ibrahim

    2012-12-01

    Full Text Available In this paper, we consider the Hyers-Ulam stability for the following fractional differential equations, in the sense ofcomplex Caputo fractional derivative defined, in the unit disk: cDßzf(z=G(f(z, cDázf(z,zf‘(z;z 0<á<1<ß<2 . Furthermore,a generalization of the admissible functions in complex Banach spaces is imposed and applications are illustrated.

  11. Guide to Rebuilding Public Sector Services in Stability Operations: A Role for the Military

    Science.gov (United States)

    2009-10-01

    equality and empower women 4. Reduce child mortality 5. Improve maternal health 6. Combat HIV/AIDS, Malaria and other diseases 7. Ensure environmental ...foundation for long-term development. Stabilization is the process by which underlying tensions that might lead to resurgence in violence and a breakdown...be an important contributing factor to ongoing fragility and the eruption of renewed conflict. This area of governance also connects to security. If

  12. Stability and Convergence of Relaxation Schemes to Hyperbolic Balance Laws via a Wave Operator

    OpenAIRE

    Miroshnikov, Alexey; Trivisa, Konstantina

    2014-01-01

    This article deals with relaxation approximations of nonlinear systems of hyperbolic balance laws. We introduce a class of relaxation schemes and establish their stability and convergence to the solution of hyperbolic balance laws before the formation of shocks, provided that we are within the framework of the compensated compactness method. Our analysis treats systems of hyperbolic balance laws with source terms satisfying a special mechanism which induces weak dissipation in the spirit of D...

  13. Simplified system for the pressure control of a Nucleo electric central of the BWR type; Sistema simplificado para el control de presion de una central Nucleoelectrica del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J. [FI-UNAM, DEPFI Campus Morelos, Jiutepec, Morelos (Mexico)

    2003-07-01

    One of the main preoccupations of the electric power generator stations is the appropriate operation of the same ones. The operators must be qualified to respond in an adequate way and to be able to take to these power stations to an optimal, sure and stable operation condition under any circumstance. The Laboratory of Analysis in Nuclear Reactors Engineering (LAIRN) of the Engineering Faculty of UNAM (Fl) in collaboration with the International Atomic Energy Agency (IAEA), it develops an interactive classroom simulator in which simulations of the phenomena which take place in a nuclear power station are executed. The classroom simulator bases its operation on specialized nuclear codes feeding interactive graphic unfolding with those that it is possible to make a monitoring, supervision and control of the behavior of the power station under any operation regime, either in normal operation, transitory events or postulated accident sequence. The development of this classroom simulator includes a modular and re configurable structure. Due to it is indispensable to count with a higher inter activity with the system it is included the simulation of the control system of the plant and inside the same, one of those more important it is the reactor pressure control system. The present work describes the conceptual design and the used methodology for the development and implementation in the simulator of a simplified model of the pressure control system for a BWR generic central. The reach of the development will allow to accomplish the necessary tests to demonstrate that this has an adequate performance according to the carried out simplifications. (Author)

  14. Effect of operational variables on nitrogen transformations in duckweed stabilization ponds

    OpenAIRE

    Caicedo Bejarano, J.R.

    2005-01-01

    There is a diversity of conventional technologies available for removal of pollutants from wastewater. Most of these technologies are aerobic alternatives with high construction cost and high energy consumption and require skilled personal for operation and maintenance. As a consequence, only countries with a high gross national product (GNP) can afford these options. Where these technologies were introduced in developing countries, in most cases these could not be operated sustainably, leadi...

  15. The expanding role of military entomologists in stability and counterinsurgency operations.

    Science.gov (United States)

    Robert, Leon L; Rankin, Steven E

    2011-01-01

    Military entomologists function as part of medical civil-military operations and are an essential combat multiplier direction supporting COIN operations. They not only directly support US and coalition military forces by performing their traditional wartime mission of protecting personnel from vector-borne and rodent-borne diseases but also enhance the legitimacy of medical services by the host nation government such as controlling diseases promulgated by food, water, vectors, and rodents. These unique COIN missions demand a new skill set required of military entomologists that are not learned from existing training courses and programs. New training opportunities must be afforded military entomologists to familiarize them with how to interact with and synergize the efforts of host nation assets, other governmental agencies, nongovernmental organizations, and international military partners. Teamwork with previously unfamiliar groups and organizations is an essential component of working in the COIN environment and can present unfamiliar tasks for entomologists. This training should start with initial entry training and be a continual process throughout a military entomologist's career. Current COIN operations require greater tactical and operational flexibility and diverse entomological expertise. The skills required for today's full spectrum medical operations are different from those of the past. Counterinsurgency medical operations demand greater agility, rapid task-switching, and the ability to adequately address unfamiliar situations and challenges.

  16. Teleoperator comfort and psychometric stability: Criteria for limiting master-controller forces of operation and feedback during telemanipulation

    Science.gov (United States)

    Wiker, Steven F.; Hershkowitz, Elaine; Zik, John

    1989-01-01

    The following question is addressed: How much force should operators exert, or experience, when operating a telemanipulator master-controller for sustained periods without encountering significant fatigue and discomfort, and without loss of stability in psychometric perception of force. The need to minimize exertion demands to avoid fatigue is diametrically opposed by the need to present a wide range of force stimuli to enhance perception of applied or reflected forces. For 104 minutes subjects repetitiously performed a series of 15 s isometric pinch grasps; controlled at 5, 15, and 25 percent of their maximum voluntary strength. Cyclic pinch grasps were separated by rest intervals of 7.5 and 15 s. Upon completion of every 10 minute period, subjects interrupted grasping activities to gage the intensity of fatigue and discomfort in the hand and forearm using a cross-modal matching technique. A series of psychometric tests were then conducted to determine accuracy and stability in the subject's perception of force experienced. Results showed that onset of sensations of discomfort and fatigue were dependent upon the magnitude of grasp force, work/rest ratio, and progression of task. Declines in force magnitude estimation slopes, indicating a reduction in force perception sensitivity, occurred with increased grasp force when work/rest ratios were greater than 1.0. Specific recommendations for avoiding discomfort and shifts in force perception, by limiting pinch grasp force required for master-controller operation and range of force reflection or work/rest ratios, are provided.

  17. Some Schrödinger-type inequalities for stabilization of discrete linear systems associated with the stationary Schrödinger operator

    Directory of Open Access Journals (Sweden)

    Zongcai Jiang

    2016-10-01

    Full Text Available Abstract By applying some Schrödinger-type inequalities developed by Huang (Int. J. Math. 27(2:1650009, 2016, we are concerned with stabilization of discrete linear systems associated with the Schrödinger operator. Our first aim is to prove a state-dependent switching law associated with the Schrödinger operator, which is based on a convex combination. Next, we derive sufficient conditions associated with the Schrödinger operator that guarantee the uniform exponential stability of the system. Finally, we propose a necessary and sufficient condition for the stability of a system with two Schrödinger subsystems.

  18. Validation of reactor noise linear stability methods by means of advanced stochastic differential equation models

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J.L., E-mail: jlcobos@iqn.upv.es [Instituto de Ingenieri' a Energetica, Universidad Politecnica de Valencia, Camino de Vera 14, Valencia 46022 (Spain); Montesinos, M.E. [Instituto de Ingenieri' a Energetica, Universidad Politecnica de Valencia, Camino de Vera 14, Valencia 46022 (Spain); Pena, J. [Departamento de Matematica Aplicada, Universidad Politecnica de Valencia, Valencia (Spain); Escriva, A.; Gonzalez, C. [Instituto de Ingenieri' a Energetica, Universidad Politecnica de Valencia, Camino de Vera 14, Valencia 46022 (Spain); Melara, J. [IBERDROLA Ingenieri' a y Construccion, Avenida Manoteras 20, Madrid 28050 (Spain)

    2011-07-15

    Highlights: > We study validation methods for stability monitoring of BWR. > We generate synthetic signals from BWR reduced order models with previously known decay ratio. > Parametric and non parametric methods are used for the prediction of the stability. > We show the optimal method for filtering in the evaluation of the decay ratio. - Abstract: The aim of this paper is to show a validation method of a stability monitor using a BWR model with multiple Wiener noise sources, of additive and multiplicative nature. This model is solved using the modern methods to integrate stochastic differential equation systems, that are based on the stochastic Ito-Taylor expansion, and developed by Kloeden and Platen (1995), Kloeden et al. (1994). The synthetic signals generated with this BWR reduced order model with multiple Wiener processes are then used to obtain what are the optimal ways of filtering the signals for the different methods to estimate the decay ration (DR) and the natural frequency ({omega}) of the system. Also, for each DR estimation method, we study what is the optimal combination of algorithms to obtain the order and coefficients of the AR model that yields the best prediction of the reactor stability parameters for a broad range of DR values.

  19. Qualification of helium measurement system for detection of fuel failures in a BWR

    Science.gov (United States)

    Larsson, I.; Sihver, L.; Loner, H.; Grundin, A.; Helmersson, J.-O.; Ledergerber, G.

    2014-05-01

    There are several methods for surveillance of fuel integrity during the operation of a boiling water reactor (BWR). The detection of fuel failures is usually performed by analysis of grab samples of off-gas and coolant activities, where a measured increased level of ionizing radiation serves as an indication of new failure or degradation of an already existing one. At some nuclear power plants the detection of fuel failures is performed by on-line nuclide specific measurements of the released fission gases in the off-gas system. However, it can be difficult to distinguish primary fuel failures from degradation of already existing failures. In this paper, a helium measuring system installed in connection to a nuclide specific measuring system to support detection of fuel failures and separate primary fuel failures from secondary ones is presented. Helium measurements provide valuable additional information to measurements of the gamma emitting fission gases for detection of primary fuel failures, since helium is used as a fill gas in the fuel rods during fabrication. The ability to detect fuel failures using helium measurements was studied by injection of helium into the feed water systems at the Forsmark nuclear power plant (NPP) in Sweden and at the nuclear power plant Leibstadt (KKL) in Switzerland. In addition, the influence of an off-gas delay line on the helium measurements was examined at KKL by injecting helium into the off-gas system. By using different injection rates, several types of fuel failures with different helium release rates were simulated. From these measurements, it was confirmed that the helium released by a failed fuel can be detected. It was also shown that the helium measurements for the detection of fuel failures should be performed at a sampling point located before any delay system. Hence, these studies showed that helium measurements can be useful to support detection of fuel failures. However, not all fuel failures which occurred at

  20. NONCONFORMING STABILIZED FINITE ELEMENT METHODS BASED ON RIESZ-REPRESENTING OPERATORS

    Institute of Scientific and Technical Information of China (English)

    DuanHuoyuan

    1999-01-01

    Following the framework of the finite element methods based on Riesz-representingoperators developed by Duan Huoyuan in 1997,through discrete Rieszonsome virtual(non-) conforming finite-dimensional subspaces,a stabilization formulation is presented for the Stokes problem by employing nonconforming elements. This formulation is uni-tormly coercive and not subject to the Babus Ka-Brezzi condition,and the resulted linearalgebraic system is positive definite with the spectral condition number O(h-2).Quasi-optimal error bounds are obtained,which is consistent with the interpolation properties of the finite elements used.

  1. Optimal Stabilization of Social Welfare under Small Variation of Operating Condition with Bifurcation Analysis

    Science.gov (United States)

    Chanda, Sandip; De, Abhinandan

    2016-12-01

    A social welfare optimization technique has been proposed in this paper with a developed state space based model and bifurcation analysis to offer substantial stability margin even in most inadvertent states of power system networks. The restoration of the power market dynamic price equilibrium has been negotiated in this paper, by forming Jacobian of the sensitivity matrix to regulate the state variables for the standardization of the quality of solution in worst possible contingencies of the network and even with co-option of intermittent renewable energy sources. The model has been tested in IEEE 30 bus system and illustrious particle swarm optimization has assisted the fusion of the proposed model and methodology.

  2. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  3. Experimental and analytical study of stability characteristics of natural circulation boiling water reactors during startup transient

    Science.gov (United States)

    Woo, Kyoungsuk

    Two-phase natural circulation loops are unstable at low pressure operating conditions. New reactor design relying on natural circulation for both normal and abnormal core cooling is susceptible to different types of flow instabilities. In contrast to forced circulation boiling water reactor (BWR), natural circulation BWR is started up without recirculation pumps. The tall chimney placed on the top of the core makes the system susceptible to flashing during low pressure start-up. In addition, the considerable saturation temperature variation may induce complicated dynamic behavior driven by thermal non-equilibrium between the liquid and steam. The thermal-hydraulic problems in two-phase natural circulation systems at low pressure and low power conditions are investigated through experimental methods. Fuel heat conduction, neutron kinetics, flow kinematics, energetics and dynamics that govern the flow behavior at low pressure, are formulated. A dimensionless analysis is introduced to obtain governing dimensionless groups which are groundwork of the system scaling. Based on the robust scaling method and start-up procedures of a typical natural circulation BWR, the simulation strategies for the transient with and without void reactivity feedback is developed. Three different heat-up rates are applied to the transient simulations to study characteristics of the stability during the start-up. Reducing heat-up rate leads to increase in the period of flashing-induced density wave oscillation and decrease in the system pressurization rate. However, reducing the heat-up rate is unable to completely prevent flashing-induced oscillations. Five characteristic regions of stability are discovered at low pressure conditions. They are stable single-phase, flashing near the separator, intermittent oscillation, sinusoidal oscillation and low subcooling stable regions. Stability maps were acquired for system pressures ranging 100 kPa to 400 kPa. According to experimental investigation

  4. Anterior transarticular screw fixation as a conventional operation for rigid stabilization

    Directory of Open Access Journals (Sweden)

    Manabu Sasaki

    2014-01-01

    Full Text Available Background: Anterior transarticular screw (ATS fixation is a useful surgical option for atlantoaxial (AA stabilization. This report presents a revised ATS method for AA fusion. Methods: A 79-year-old male presented with AA instability attributed both to an old odontoid fracture and severe degeneration of the lateral atlantoaxial joints (LAAJs. ATS fixation was performed through the conventional anterior cervical approach. The longest screw trajectories were planned preoperatively using multiplanar reconstruction computed tomography (CT scans, with entry points of the screws situated at the midpoint on the inferior border of the axial body. The surgical exposure was limited to opening at the entry points alone. Our retractor of choice was the Cusco speculum; it sufficiently secured space for utilizing the required instruments for screw placement while offering sufficient protection of soft tissues. Cannulated full-threaded bicortical screws stabilized the LAAJs. Screw insertion required a significant amount of coronal angulation up to the superior articular process of the atlas under open-mouth and lateral fluoroscopy image guidance. After ATS fixation, bone grafting was performed between the posterior laminae of the axis and the atlas through a conventional posterior approach. Results: Bony fusion between the atlas and the axis was confirmed radiographically. Arthrodesis of the LAAJs occurred despite no bone grafting. Conclusions: Rigid fixation of the LAAJs was obtained by our ATS technique, indicating that it is an alternative method for AA fixation when posterior rigid internal fixation is not applicable.

  5. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  6. Effect of operational variables on nitrogen transformations in duckweed stabilization ponds

    NARCIS (Netherlands)

    Caicedo Bejarano, J.R.

    2005-01-01

    There is a diversity of conventional technologies available for removal of pollutants from wastewater. Most of these technologies are aerobic alternatives with high construction cost and high energy consumption and require skilled personal for operation and maintenance.

  7. Reconstruction and Stability Operations: Lead, Follow, or Get Out of the Way

    Science.gov (United States)

    2009-05-04

    But Key to Rebuilding Efforts,” (FDCH Regulatory Intelligence Database , Department of Defense) ebscohost (Region Group 12: 12 February 2009). 45...47 “Projects Prove Difficult, But Key to Rebuilding Efforts,” (FDCH Regulatory Intelligence Database , Department of Defense) ebscohost (Region...12 February 2009. FDCH Regulatory Intelligence Database . Forward Operating Base Gardez, Afghanistan: EBSCOhost . [12 February 2009]. Amato

  8. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm{sup 2}); Comportamiento a la fractura de un acero inoxidable AISI 304 sensibilizado en condiciones de reactor BWR (288 grados Centigrados y 80 Kg/cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm{sup 2}). (Author)

  9. Actinides record, power calculations and activity for present isotopes in the spent fuel of a BWR; Historial de actinidos y calculos de potencia y actividad para isotopos presentes en el combustible gastado de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Lucatero, M. A., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The administration of spent fuel is one of the more important stages of the nuclear fuel cycle, and this has become a problem of supreme importance in countries that possess nuclear reactors. Due to this in this work, the study on the actinides record and present fission products to the discharge of the irradiated fuel in a light water reactor type BWR is shown, to quantify the power and activity that emit to the discharge and during the cooling time. The analysis was realized on a fuel assembly type 10 x 10 with an enrichment average of 3.69 wt % in U-235 and the assembly simulation assumes four cycles of operation of 18 months each one and presents an exposition of 47 G Wd/Tm to the discharge. The module OrigenArp of the Scale 6 code is the computation tool used for the assembly simulation and to obtain the results on the actinides record presents to the fuel discharge. The study covers the following points: a) Obtaining of the plutonium vector used in the fuel production of mixed oxides, and b) Power calculation and activity for present actinides to the discharge. The results presented in this work, correspond at the same time immediate of discharge (0 years) and to a cooling stage in the irradiated fuel pool (5 years). (Author)

  10. Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff`s basis for issuing GL 94-03, as well as the staff`s assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date.

  11. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    Energy Technology Data Exchange (ETDEWEB)

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data.

  12. Polymer Nanodot-Hybridized Alkyl Silicon Oxide Nanostructures for Organic Memory Transistors with Outstanding High-Temperature Operation Stability

    Science.gov (United States)

    Lee, Chulyeon; Seo, Jooyeok; Kim, Jeongnam; Jeong, Jaehoon; Han, Hyemi; Kim, Hwajeong; Kim, Youngkyoo

    2016-01-01

    Organic memory devices (OMDs) are becoming more important as a core component in flexible electronics era because of their huge potentials for ultrathin, lightweight and flexible plastic memory modules. In particular, transistor-type OMDs (TOMDs) have been gradually spotlighted due to their structural advantages possessing both memory and driving functions in single devices. Although a variety of TOMDs have been developed by introducing various materials, less attention has been paid to the stable operation at high temperatures. Here we demonstrate that the polymer nanodot-embedded alkyl silicon oxide (ASiO) hybrid materials, which are prepared by sol-gel and thermal cross-linking reactions between poly(4-vinylphenol) (PVP) and vinyltriethoxysilane, can deliver low-voltage (1~5 V) TOMDs with outstanding operation stability (>4700 cycles) at high temperatures (150 °C). The efficient low-voltage memory function is enabled by the embedded PVP nanodots with particular lattice nanostructures, while the high thermal stability is achieved by the cross-linked ASiO network structures. PMID:27703187

  13. Polymer Nanodot-Hybridized Alkyl Silicon Oxide Nanostructures for Organic Memory Transistors with Outstanding High-Temperature Operation Stability

    Science.gov (United States)

    Lee, Chulyeon; Seo, Jooyeok; Kim, Jeongnam; Jeong, Jaehoon; Han, Hyemi; Kim, Hwajeong; Kim, Youngkyoo

    2016-10-01

    Organic memory devices (OMDs) are becoming more important as a core component in flexible electronics era because of their huge potentials for ultrathin, lightweight and flexible plastic memory modules. In particular, transistor-type OMDs (TOMDs) have been gradually spotlighted due to their structural advantages possessing both memory and driving functions in single devices. Although a variety of TOMDs have been developed by introducing various materials, less attention has been paid to the stable operation at high temperatures. Here we demonstrate that the polymer nanodot-embedded alkyl silicon oxide (ASiO) hybrid materials, which are prepared by sol-gel and thermal cross-linking reactions between poly(4-vinylphenol) (PVP) and vinyltriethoxysilane, can deliver low-voltage (1~5 V) TOMDs with outstanding operation stability (>4700 cycles) at high temperatures (150 °C). The efficient low-voltage memory function is enabled by the embedded PVP nanodots with particular lattice nanostructures, while the high thermal stability is achieved by the cross-linked ASiO network structures.

  14. First Time Go: Creating Capacity for Enduring Stability in Post-Operational Environments

    Science.gov (United States)

    2013-05-23

    Institute of Peace Press, 1997), 95-95, 89-92. 53Dobbins, “Lessons,” in Fukuyama, 226; Zoltan Barany , The Soldier and the Changing State: 13...METLs identify key tasks in which units must become proficient. Units test METL competencies as part of their regular training activities. These...as Chairman of the Joint Chiefs of Staff. Combatant commanders followed this doctrine as they began operations in Iraq. The policy was tested in

  15. The Fatal Five? Five Factors That Enhance Effectiveness of Stability Operations

    Science.gov (United States)

    2014-05-22

    government, began to collapse. Months of harassment of U.S. military personnel stationed in Panama, as well as accusations against Noriega (made by...federal grand juries indicted Noriega on a number of criminal charges, resulting in a sharp increase in harassment of U.S. military personnel...additional mission of removing Noriega from power and extraditing him to the United States.108 Upon execution, Operation Just Cause (originally

  16. SOLLIMS Sampler: Targeting Peace & Stability Operations Lessons & Best Practices. Volume 3, Issue 2

    Science.gov (United States)

    2012-04-01

    Dr. Roshan, a licensed veterinarian who is part of a team directed by the Kunar provincial veterinarian . (U.S. Air Force photo by Captain Peter...the military leadership of a Provincial Reconstruction Team (PRT) located in Panjshir had planned a program to bring American veterinarians to...Table of Contents | Quick Look | Contact PKSOI Page 27 of 42 USNS Comfort, equipped with surgical operating teams and orthopedic

  17. Long-Term Stability of Underground Operated CZT Detectors Based on the Analysis of Intrinsic $^{113}$Cd \\beta$^{-}$-Decay

    CERN Document Server

    Ebert, J; Gehre, D; Hagner, C; Heidrich, N; Klingenberg, R; Kroeninger, K; Nitsch, C; Oldorf, C; Quante, T; Rajek, S; Rebber, H; Rohatsch, K; Tebruegge, J; Temminghoff, R; Theinert, R; Timm, J; Wonsak, B; Zatschler, S; Zuber, K

    2015-01-01

    The COBRA collaboration operates a demonstrator setup at the underground facility LNGS (Laboratori Nazionali del Gran Sasso, located in Italy) to prove the technological capabilities of this concept for the search for neutrinoless double beta-decay. The setup consists of 64 $(1\\times\\!1\\times\\!1)$ cm$^{3}$ CZT detectors in CPG configuration. One purpose of this demonstrator is to test if reliable long-term operation of CZT-CPG detectors in such a setup is possible. The demonstrator has been operated under ultra low-background conditions since more than three years and collected data corresponding to an exposure of 218 kg$\\cdot$days. The presented study focuses on the long-term stability of CZT detectors by analyzing the intrinsic, fourfold forbidden non-unique $^{113}$Cd single beta-decay. It can be shown that CZT detectors can be operated stably for long periods of time and that the $^{113}$Cd single beta-decay can be used as an internal monitor of the detector performance during the runtime of the experimen...

  18. Development of an Input Model to MELCOR 1.8.5 for the Oskarshamn 3 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Lars [Lentek, Nykoeping (Sweden)

    2006-05-15

    An input model has been prepared to the code MELCOR 1.8.5 for the Swedish Oskarshamn 3 Boiling Water Reactor (O3). This report describes the modelling work and the various files which comprise the input deck. Input data are mainly based on original drawings and system descriptions made available by courtesy of OKG AB. Comparison and check of some primary system data were made against an O3 input file to the SCDAP/RELAP5 code that was used in the SARA project. Useful information was also obtained from the FSAR (Final Safety Analysis Report) for O3 and the SKI report '2003 Stoerningshandboken BWR'. The input models the O3 reactor at its current state with the operating power of 3300 MW{sub th}. One aim with this work is that the MELCOR input could also be used for power upgrading studies. All fuel assemblies are thus assumed to consist of the new Westinghouse-Atom's SVEA-96 Optima2 fuel. MELCOR is a severe accident code developed by Sandia National Laboratory under contract from the U.S. Nuclear Regulatory Commission (NRC). MELCOR is a successor to STCP (Source Term Code Package) and has thus a long evolutionary history. The input described here is adapted to the latest version 1.8.5 available when the work began. It was released the year 2000, but a new version 1.8.6 was distributed recently. Conversion to the new version is recommended. (During the writing of this report still another code version, MELCOR 2.0, has been announced to be released within short.) In version 1.8.5 there is an option to describe the accident progression in the lower plenum and the melt-through of the reactor vessel bottom in more detail by use of the Bottom Head (BH) package developed by Oak Ridge National Laboratory especially for BWRs. This is in addition to the ordinary MELCOR COR package. Since problems arose running with the BH input two versions of the O3 input deck were produced, a NONBH and a BH deck. The BH package is no longer a separate package in the new 1

  19. Analyses of containment source term of BWR5 considering iodine chemistry suppression pool with THALES-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Jun; Moriyama, Kiyofumi [Japan Atomic Energy Agency, Ibaraki (Japan)

    2009-05-15

    After JCO criticality accident in 1999, recognized the importance of PSA application research for emergency planning and basic technical study supporting decision making in protective actions. In order to evaluate containment source term in the late phase SA, coupling of severe accident analysis code THALES-2 and kinetics of iodine chemistry code Kiche was done. And containment source term analyses were performed a typical accident sequence TQUV of BWR5/Mark-II. The lower the pH in the pool was, the more fraction of iodine were released to gas phase, as was in agreement with the known tendency. Total release fractions of all iodine species to gas phase at 40 hr were 0.1[-](pH=5), 0.01[-](pH=7), 4x10{sup -4}[-] (pH=9). I{sub 2} was dominant in released iodine to gas phase and most of released I{sub 2} was adsorbed to the wall. As the operation of the containment spray, the release of iodine tot the gas phase was enhanced due to the break of a steady state by the circulation in the containment. In future, JAEA will perform containment source term analyses for extensive accident sequences with consideration of iodine chemistry.

  20. Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2002-07-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

  1. Impact of ideal MHD stability limits on high-beta hybrid operation

    Science.gov (United States)

    Piovesan, P.; Igochine, V.; Turco, F.; Ryan, D. A.; Cianciosa, M. R.; Liu, Y. Q.; Marrelli, L.; Terranova, D.; Wilcox, R. S.; Wingen, A.; Angioni, C.; Bock, A.; Chrystal, C.; Classen, I.; Dunne, M.; Ferraro, N. M.; Fischer, R.; Gude, A.; Holcomb, C. T.; Lebschy, A.; Luce, T. C.; Maraschek, M.; McDermott, R.; Odstrčil, T.; Paz-Soldan, C.; Reich, M.; Sertoli, M.; Suttrop, W.; Taylor, N. Z.; Weiland, M.; Willensdorfer, M.; The ASDEX Upgrade Team; The DIII-D Team; The EUROfusion MST1 Team

    2017-01-01

    The hybrid scenario is a candidate for stationary high-fusion gain tokamak operation in ITER and DEMO. To obtain such performance, the energy confinement and the normalized pressure {βN} must be maximized, which requires operating near or above ideal MHD no-wall limits. New experimental findings show how these limits can affect hybrid operation. Even if hybrids are mainly limited by tearing modes, proximity to the no-wall limit leads to 3D field amplification that affects plasma profiles, e.g. rotation braking is observed in ASDEX Upgrade throughout the plasma and peaks in the core. As a result, even the small ASDEX Upgrade error fields are amplified and their effects become visible. To quantify such effects, ASDEX Upgrade measured the response to 3D fields applied by 8× 2 non-axisymmetric coils as {βN} approaches the no-wall limit. The full n  =  1 response profile and poloidal structure were measured by a suite of diagnostics and compared with linear MHD simulations, revealing a characteristic feature of hybrids: the n  =  1 response is due to a global, marginally-stable n  =  1 kink characterized by a large m  =  1, n  =  1 core harmonic due to q min being just above 1. A helical core distortion of a few cm forms and affects various core quantities, including plasma rotation, electron and ion temperature, and intrinsic W density. In similar experiments, DIII-D also measured the effect of this helical core on the internal current profile, providing information useful to understanding of the physics of magnetic flux pumping, i.e. anomalous current redistribution by MHD modes that keeps {{q}\\text{min}}>1 . Thanks to flux pumping, a broad current profile is maintained in DIII-D even with large on-axis current drive, enabling fully non-inductive operation at high {βN} up to 3.5-4.

  2. Step Sizes for Strong Stability Preservation with Downwind-Biased Operators

    KAUST Repository

    Ketcheson, David I.

    2011-08-04

    Strong stability preserving (SSP) integrators for initial value ODEs preserve temporal monotonicity solution properties in arbitrary norms. All existing SSP methods, including implicit methods, either require small step sizes or achieve only first order accuracy. It is possible to achieve more relaxed step size restrictions in the discretization of hyperbolic PDEs through the use of both upwind- and downwind-biased semidiscretizations. We investigate bounds on the maximum SSP step size for methods that include negative coefficients and downwind-biased semi-discretizations. We prove that the downwind SSP coefficient for linear multistep methods of order greater than one is at most equal to two, while the downwind SSP coefficient for explicit Runge–Kutta methods is at most equal to the number of stages of the method. In contrast, the maximal downwind SSP coefficient for second order Runge–Kutta methods is shown to be unbounded. We present a class of such methods with arbitrarily large SSP coefficient and demonstrate that they achieve second order accuracy for large CFL number.

  3. Application of the FFTBM method and the power relative contribution to the discharge transitory of the recirculation pumps of a BWR; Aplicacion del metodo FFTBM y de la contribucion relativa de potencia al transitorio de disparo de las bombas de recirculacion de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J.; Fuentes M, L., E-mail: rogelio.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In this work was realized the simulation of the discharge transitory of both recirculation pumps of a BWR with the Simulate-3K code. This type of transitory is used in the stability analyses for the licensing of the fuel reload. An analysis of the precision of the simulation is also presented, using the FFTBM method jointly with the power relative contribution. This way, instead of determining the total precision of the calculation, a weighed precision is obtained by the contribution of each relevant parameter of the transitory. The results show that the precision of the simulation is acceptable due to the small magnitude of the merit figure of the width total average. The error in the merit figure comes mainly from the parameters total flow in the core and temperature of the fuel in the core. (Author)

  4. Stability of Flip and Exchange Symmetric Entangled State Classes under Invertible Local Operations

    CERN Document Server

    Gedik, Zafer

    2009-01-01

    Flip and exchange symmetric (FES) many-qubit states, which can be obtained from a state with the same symmetries by means of invertible local operations (ILO), constitute a one-parameter family of curves in the Hilbert space. Eigenstates of FES ILOs correspond to vectors that cannot be transformed to other FES states. This means equivalence classes of states under ILO can be determined in a systematic way for an arbitrary number of qubits. More important, for entangled states, at the boundaries of neighboring equivalence classes, one can show that when the fidelity between the final state after an ILO and a state of the neighboring class approaches unity, probability of success decreases to zero. Therefore, the classes are stable under ILOs.

  5. Stability of flip and exchange symmetric entangled state classes under invertible local operations

    Science.gov (United States)

    Gedik, Z.

    2011-01-01

    Flip and exchange symmetric (FES) many-qubit states, which can be obtained from a state with the same symmetries by means of invertible local operations (ILO), constitute a set of curves in the Hilbert space. Eigenstates of FES ILOs correspond to vectors that cannot be transformed to other FES states. This means equivalence classes of states under ILO can be determined in a systematic way for an arbitrary number of qubits. More important, for entangled states, at the boundaries of neighboring equivalence classes, one can show that when the fidelity between the final state after an ILO and a state of the neighboring class approaches unity, the probability of success decreases to zero. Therefore, the classes are stable under ILOs.

  6. A roll-to-roll process to flexible polymer solar cells: model studies, manufacture and operational stability studies

    DEFF Research Database (Denmark)

    Krebs, Frederik C; Gevorgyan, Suren; Alstrup, Jan

    2009-01-01

    An inverted polymer solar cell geometry comprising a total of five layers was optimized using laboratory scale cells and the operational stability was studied under model atmospheres. The device geometry was substrate-ITO-ZnO-(active layer)-PEDOT:PSS-silver with P3HT-PCBM as the active layer...... efficiency of 2.7% (1000 W m-2, AM1.5G, 72 ± 2 °C). The devices were operated under illumination while being subjected to different atmospheres to identify the dominant modes of degradation. Dry nitrogen (99.999%), dry oxygen (99.5%), humid nitrogen (90 ± 5% relative humidity) and ambient atmosphere (20......% oxygen, 20 ± 5% relative humidity) were employed and both device types were found to be stable in a nitrogen atmosphere during the test period of 200 hours. The devices with a normal geometry where an aluminium electrode is employed gave stable operation in dry oxygen but did not give stable device...

  7. Effects of Environmental Factors on the Disinfection Performance of a Wastewater Stabilization Pond Operated in a Temperate Climate

    Directory of Open Access Journals (Sweden)

    Lei Liu

    2015-12-01

    Full Text Available Treatment in a wastewater stabilization pond (WSP relies on natural purification processes, which can be sensitive to both location and climate. This study investigated the effects of three environmental factors, pH, dissolved oxygen (DO and temperature, on disinfection efficiency in a WSP system consisting of three facultative cells, and operated in a temperate climate region, in Eastern Ontario, Canada. Indicator organism (Escherichia coli (E. coli removal in WSP systems is driven by a combination of different factors. Elevated pH and DO concentrations, which are attributed to the presence of algae, are important factors for effective disinfection. Therefore, the presence of algae in natural wastewater treatment systems can contribute appreciably to disinfection. Consequently, based on algal concentrations, removal efficiencies of pathogenic microorganisms during wastewater treatment over the course of a year can be highly variable, where higher removal efficiencies would be expected in summer and fall seasons.

  8. 78 FR 22563 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

    Science.gov (United States)

    2013-04-16

    ... reverse osmosis system during normal plant operation to purify the water in the borated water storage... Selected Required Action End States for BWR Plants.'' Additionally, the proposed amendment would modify the... train of equipment or a restriction on a plant operational parameter, unless otherwise stated in...

  9. Geo-mechanical Model Testing for Stability of Underground Gas Storage in Halite During the Operational Period

    Science.gov (United States)

    Chen, Xuguang; Zhang, Qiangyong; Li, Shucai; Liu, Dejun

    2016-07-01

    A 3D geo-mechanical model test is conducted to study the stability of underground gas storage in halite, modeled after the Jintan gas storage constructed in bedded salt rock in China. A testing apparatus is developed to generate long-term stable trapezoid geostresses onto the model cavity, corresponding to the actual gas storage cavern. The time-depending character of the material is simulated using a rheological material, which was tested using a self-developed apparatus. The model cavern is built using an ellipsoid wooden mold divided into small blocks which are assembled and placed into the designed position during the model construction. They are then pulled out one by one to form the cavern. The ellipsoid cavern wall is then lined within a latex balloon. Gas is injected into the cavity and extracted to simulate the operational process of gas injection and recovery. Optical sensors embedded into the model to measure the displacement around the cavity showed that the largest deformation occurs in the middle section of the cavity. The deformation rate increases with increasing gas pressure. At 11 MPa the cavity is in equilibrium with the geostress. The pressure is highest during the gas recovery stages, indicating that gas recovery can threaten the cavern's operational stability, while high gas injection causes rock mass compression and deformation outward from the cavern. The deformation is the combination of cavern convergence and gas-induced rebound which leads to tensile and compression during gas injection and recovery. Hence, the fatigue properties of salt rock should be studied further.

  10. PVT modeling of reservoir fluids using PC-SAFT EoS and Soave-BWR EoS

    DEFF Research Database (Denmark)

    Yan, Wei; Varzandeh, Farhad; Stenby, Erling Halfdan

    2015-01-01

    non-cubic EoS models, such as the Perturbed Chain Statistical Associating Fluid Theory (PC-SAFT) EoS and the Soave modified Benedict-Webb-Rubin (Soave-BWR) EoS, may partly replace the roles of these classical cubic models in the upstream oil industry. Here, we attempt to make a comparative study...... of non-cubic models (PC-SAFT and Soave-BWR) and cubic models (SRK and PR) in several important aspects related to PVT modeling of reservoir fluids, including density description for typical pure components in reservoir fluids, description of binary VLE, prediction of multicomponent phase envelopes...... and Soave-BWR. For PC-SAFT, new correlations for estimating its model parameters in heptanes plus are developed. The results reveal that the non-cubic models are clearly advantageous in density calculation of pure components. For binary VLE and multicomponent phase envelopes, the results are similar...

  11. Analysis of high fidelity of a BWR fuel element with COBRA-TF/PARCS codes and TRACE; Analisis de Alta Fidelidad de un Elemento Combustible BWR con los codigos COBRA-TF/PARCS y TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Solar, A.; Concejal, A.; Melara, J.; Albendea, M.

    2013-07-01

    It has been modeled a 10 x 10 BWR fuel element, containing 91 fuel rods (81 of 10 partial length and total length) and a great water bar of square section in the central part of it. Such fuel element has been modeled in detail: at the level of sub-channel code COBRA-TF and using parametric models for fuel elements BWR that owns the plant code TRACE. Has been an exercise in comparison of the results obtained by both codes in the simulation of a stationary and a small transient flow injection, highlighting the differences observed.

  12. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2; Analisis conceptual de un modelo preliminar para el estudio de la inestabilidad en la operacion normal de un reactor de circulacion natural tipo ESBWR, usando Relap5/Mod 3.2

    Energy Technology Data Exchange (ETDEWEB)

    Ojeda S, J.; Morales S, J.; Chavez M, C. [UNAM, Facultad de Ingenieria, Circuito Exterior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: j.os.ojeda@hotmail.com

    2009-10-15

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  13. Reactor scram experience for shutdown system reliability analysis. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.

    1976-06-01

    Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.

  14. Component failures that lead to reactor scrams. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Burns, E. T.; Wilson, R. J.; Lim, E. Y.

    1980-04-01

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation.

  15. Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

    OpenAIRE

    2004-01-01

    Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8times8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding...

  16. High operational and environmental stability of high-mobility conjugated polymer field-effect transistors through the use of molecular additives

    Science.gov (United States)

    Nikolka, Mark; Nasrallah, Iyad; Rose, Bradley; Ravva, Mahesh Kumar; Broch, Katharina; Sadhanala, Aditya; Harkin, David; Charmet, Jerome; Hurhangee, Michael; Brown, Adam; Illig, Steffen; Too, Patrick; Jongman, Jan; McCulloch, Iain; Bredas, Jean-Luc; Sirringhaus, Henning

    2016-12-01

    Due to their low-temperature processing properties and inherent mechanical flexibility, conjugated polymer field-effect transistors (FETs) are promising candidates for enabling flexible electronic circuits and displays. Much progress has been made on materials performance; however, there remain significant concerns about operational and environmental stability, particularly in the context of applications that require a very high level of threshold voltage stability, such as active-matrix addressing of organic light-emitting diode displays. Here, we investigate the physical mechanisms behind operational and environmental degradation of high-mobility, p-type polymer FETs and demonstrate an effective route to improve device stability. We show that water incorporated in nanometre-sized voids within the polymer microstructure is the key factor in charge trapping and device degradation. By inserting molecular additives that displace water from these voids, it is possible to increase the stability as well as uniformity to a high level sufficient for demanding industrial applications.

  17. High operational and environmental stability of high-mobility conjugated polymer field-effect transistors through the use of molecular additives

    KAUST Repository

    Nikolka, Mark

    2016-12-12

    Due to their low-temperature processing properties and inherent mechanical flexibility, conjugated polymer field-effect transistors (FETs) are promising candidates for enabling flexible electronic circuits and displays. Much progress has been made on materials performance; however, there remain significant concerns about operational and environmental stability, particularly in the context of applications that require a very high level of threshold voltage stability, such as active-matrix addressing of organic light-emitting diode displays. Here, we investigate the physical mechanisms behind operational and environmental degradation of high-mobility, p-type polymer FETs and demonstrate an effective route to improve device stability. We show that water incorporated in nanometre-sized voids within the polymer microstructure is the key factor in charge trapping and device degradation. By inserting molecular additives that displace water from these voids, it is possible to increase the stability as well as uniformity to a high level sufficient for demanding industrial applications.

  18. Obtention control bars patterns for a BWR using Tabo search; Obtencion de patrones de barras de control para un BWR usando busqueda Tabu

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, A.; Ortiz, J.J.; Alonso, G. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Ocoyoacac, Estado de Mexico 52045 (Mexico); Morales, L.B. [UNAM, IIMAS, Ciudad Universitaria, D. F. 04510 (Mexico); Valle, E. del [IPN, ESFM, Unidad Profesional ' Adolfo Lopez Mateos' , Col. Lindavista 07738, D. F. (Mexico)]. e-mail: jacm@nuclear.inin.mx

    2004-07-01

    The obtained results when implementing the technique of tabu search, for to optimize patterns of control bars in a BWR type reactor, using the CM-PRESTO code are presented. The patterns of control bars were obtained for the designs of fuel reloads obtained in a previous work, using the same technique. The obtained results correspond to a cycle of 18 months using 112 fresh fuels enriched at the 3.53 of U-235. The used technique of tabu search, prohibits recently visited movements, in the position that correspond to the axial positions of the control bars, additionally the tiempo{sub t}abu matrix is used for to manage a size of variable tabu list and the objective function is punished with the frequency of the forbidden movements. The obtained patterns of control bars improve the longitude of the cycle with regard to the reference values and they complete the restrictions of safety. (Author)

  19. Analysis of assemblies exchange in the core of a reactor BWR; Analisis del intercambio de ensambles en el nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kauil U, J. S. [Universidad Autonoma de Yucatan, Facultad de Ingenieria, Av. Industrias no contaminantes por Anillo Periferico Norte s/n, Apdo. Postal 150 Cordemex, Merida, Yucatan (Mexico); Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia del Cueto, R., E-mail: san_dino@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The performance of the core of a boiling water reactor (BWR) was evaluated when two assemblies are exchanged during the fuel reload in erroneous way. All with the purpose of analyzing the value of the neutrons effective multiplication factor and the thermal limits for an exchange of assemblies. In their realization the mentioned study was based in a transition cycle of the Unit 1 of the nuclear power plant of Laguna Verde. The obtained results demonstrate that when carrying out an exchange between two fuel assemblies in erroneous way, with regard to the original reload, the changes in the neutrons effective multiplication factor do not present a serious problem, unless the exchange has been carried out among a very burnt assembly with one fresh, where this last is taken to the periphery. (Author)

  20. Crack growth rate in core shroud horizontal welds using two models for a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis Juárez, C.R., E-mail: carlos.arganis@inin.gob.mx; Hernández Callejas, R.; Medina Almazán, A.L.

    2015-05-15

    Highlights: • Two models were used to predict SCC growth rate in a core shroud of a BWR. • A weld residual stress distribution with 30% stress relaxation by neutron was used. • Agreement is shown between the measurements of SCC growth rate and the predictions. • Slip–oxidation model is better at low fluences and empirical model at high fluences. - Abstract: An empirical crack growth rate correlation model and a predictive model based on the slip–oxidation mechanism for Stress Corrosion Cracking (SCC) were used to calculate the crack growth rate in a BWR core shroud. In this study, the crack growth rate was calculated by accounting for the environmental factors related to aqueous environment, neutron irradiation to high fluence and the complex residual stress conditions resulting from welding. In estimating the SCC behavior the crack growth measurements data from a Boiling Water Reactor (BWR) plant are referred to, and the stress intensity factor vs crack depth throughout thickness is calculated using a generic weld residual stress distribution for a core shroud, with a 30% stress relaxation induced by neutron irradiation. Quantitative agreement is shown between the measurements of SCC growth rate and the predictions of the slip–oxidation mechanism model for relatively low fluences (5 × 10{sup 24} n/m{sup 2}), and the empirical model predicted better the SCC growth rate than the slip–oxidation model for high fluences (>1 × 10{sup 25} n/m{sup 2}). The relevance of the models predictions for SCC growth rate behavior depends on knowing the model parameters.

  1. Development of Sequential Calculating Type Autonomous Stabilizing Controller for Maintaining Transient Stability in case of Backup Relay Operation in Load System

    Science.gov (United States)

    Kobayashi, Ryoji; Maekawa, Kazuo; Shimomura, Kimihiko; Sasaki, Takashi; Kowada, Yasuyuki; Maeda, Toru

    A new Autonomous Stabilizing Controller (ASC) has been developed, which is based on the concept of using only locall information measured at the substation where ASC is installed. ASC prevents the power system from wide-area blackout by shedding generators optimally when a serious fault occurs in 154kV-or-below-voltage power system and is cleared by backup relay. ASC consists of three functions, a function of detecting faults and judging of starting stability calculations (The first function), a function of calculating power system stability (The second function), and a function of selecting optimal generators to be shed (The third function). The first function composes a signal for starting stability calculations when a fault cleared by backup relay is detected. According to the starting signal composed by the first function, the second function executes stability calculations based on equal area criterion. If the result of stability calculations is instable, the third function selects a generator to be shed by considering not only the stabilizing effect but also the quickly restoration after shedding. This paper presents total algorithms applied in ASC.

  2. The effect of metal-buffer bilayer drain/source electrodes on the operational stability of the organic field effect transistors

    Energy Technology Data Exchange (ETDEWEB)

    Karimi-Alavijeh, H.R., E-mail: h.karimi@eng.ui.ac.ir [Department of Electrical Engineering, University of Isfahan, Isfahan (Iran, Islamic Republic of); Ehsani, A. [Department of Electrical and Avionics Engineering, Malek-Ashtar University of Technology, Isfahan (Iran, Islamic Republic of)

    2015-09-01

    In this paper, we have investigated experimentally the effect of different drain/source (D/S) electrodes and charge injection buffer layers on the electrical properties and operational stability of a stilbene organic field effect transistor (OFET). The results show that the organic buffer layer of copper phthalocyanine (CuPc) considerably improves the electrical properties of the transistors, but has a negligible effect on their temporal behavior. On the other hand, inorganic metal-oxide buffer layer of molybdenum oxide (MoO{sub 3}) drastically changes both the electrical properties and operational stability. The functionalities of this metal-oxide tightly depend on the properties of the D/S metallic electrodes. OFETs with Al/MoO{sub 3} as the bilayer D/S electrodes have the best electrical properties: field effect mobility μ{sub eff} = 0.32 cm{sup 2} V{sup −1} s{sup −1} and threshold voltage V{sub TH} = − 5 V and the transistors with Ag/MoO{sub 3} have the longest operational stability. It was concluded that the chemical stability of the metal/metal-oxide or metal/organic interfaces of the bilayer D/S electrodes determine the operational stability of the OFETs. - Highlights: • The effect of buffer layers on the performance of the stilbene OFETs has been investigated. • Inorganic buffer layer improved the electrical and temporal behaviors simultaneously. • Organic buffer layer only changes the electrical properties. • Chemical stability of the interfaces determines the operational stability of the transistor.

  3. A compact iodine-laser operating at 531 nm with stability at the 10$^{-12}$ level and using a coin-sized laser module

    CERN Document Server

    Kobayashi, Takumi; Hosaka, Kazumoto; Inaba, Hajime; Okubo, Sho; Tanabe, Takehiko; Yasuda, Masami; Onae, Atsushi; Hong, Feng-Lei

    2015-01-01

    We demonstrate a compact iodine-stabilized laser operating at 531 nm using a coin-sized light source consisting of a 1062-nm distributed-feedback diode laser and a frequency-doubling element. A hyperfine transition of molecular iodine is observed using the light source with saturated absorption spectroscopy. The light source is frequency stabilized to the observed iodine transition and achieves frequency stability at the 10$^{-12}$ level. The absolute frequency of the compact laser stabilized to the $a_{1}$ hyperfine component of the $R(36)32-0$ transition is determined as $564\\,074\\,632\\,419(8)$ kHz with a relative uncertainty of $1.4\\times10^{-11}$. The iodine-stabilized laser can be used for various applications including interferometric measurements.

  4. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant; Implementacion digital, simulacion y pruebas en MATLAB de los modelos de la linea de vapor, las turbinas y el regulador de presion de una central Nucleoelectrica tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: andyskamx@yahoo.com.mx

    2004-07-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  5. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  6. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  7. LOCA steam condensation loads in BWR Mark II pressure suppression containment system

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Takeshita, I.; Shiba, M.

    1987-06-01

    Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates (approx. = 30 kg/m/sup 2/.s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.

  8. LOCA air-injection loads in BWR Mark II pressure suppression containment systems

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Shiba, M. (Japan Atomic Energy Research Inst., Tokai, Ibaraki); Namatame, K. (Institute of Nuclear Safety, Tokyo (Japan))

    1984-02-01

    Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.

  9. Effect of non-heterogeneous wetwell boundaries on pressure suppression system response. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    McCauley, E.W.; Holman, G.S.; Namatame, K.; Kukita, Y.; Shiba, M.

    1980-08-29

    The Full-Scale Mark II CRT (Containment Response Test) Program is in progress at the Tokai-Mura Establishment of the Japan Atomic Energy Research Institute (JAERI). The primary objective of the on-going CRT Program is to provide a data base for evaluation of the pressure suppression pool (wetwell) hydrodynamic loads associated with a postulated loss-of-coolant accident (LOCA) in the BWR Mark II containment system. The test facility is 1/18 of full scale in volume and has a wetwell which is a full-scale geometric replica of one 20/sup 0/-sector of a reference 1100MWe Mark II.

  10. TRACE code validation for BWR spray cooling injection based on GOTA facility experiments

    Energy Technology Data Exchange (ETDEWEB)

    Racca, S. [San Piero a Grado Nuclear Research Group (GRNSPG), Pisa (Italy); Kozlowski, T. [Royal Inst. of Tech., Stockholm (Sweden)

    2011-07-01

    Best estimate codes have been used in the past thirty years for the design, licensing and safety of NPP. Nevertheless, large efforts are necessary for the qualification and the assessment of such codes. The aim of this work is to study the main phenomena involved in the emergency spray cooling injection in a Swedish designed BWR. For this purpose, data from the Swedish separate effect test facility GOTA have been simulated using TRACE version 5.0 Patch 2. Furthermore, uncertainty calculations have been performed with the propagation of input errors method and the identification of the input parameters that mostly influence the peak cladding temperature has been performed. (author)

  11. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde; Estudio de ruido ambiental en una planta BWR como la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  12. Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR; Determinacion de esfuerzos originados por fluctuacion de carga acustica en los secadores de vapor de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A., E-mail: javcuami26@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

  13. Failure Mechanisms and Color Stability in Light-Emitting Diodes during Operation in High- Temperature Environments in Presence of Contamination

    Energy Technology Data Exchange (ETDEWEB)

    Lall, Pradeep; Zhang, Hao; Davis, J Lynn

    2015-05-26

    The energy efficiency of light-emitting diode (LED) technology compared to incandescent light bulbs has triggered an increased focus on solid state luminaries for a variety of lighting applications. Solid-state lighting (SSL) utilizes LEDs, for illumination through the process of electroluminescence instead of heating a wire filament as seen with traditional lighting. The fundamental differences in the construction of LED and the incandescent lamp results in different failure modes including lumen degradation, chromaticity shift and drift in the correlated color temperature. The use of LED-based products for safety-critical and harsh environment applications necessitates the characterization of the failure mechanisms and modes. In this paper, failure mechanisms and color stability has been studied for commercially available vertical structured thin film LED (VLED) under harsh environment conditions with and without the presence of contaminants. The VLED used for the study was mounted on a ceramic starboard in order to connect it to the current source. Contamination sources studied include operation in the vicinity of vulcanized rubber and adhesive epoxies in the presence of temperature and humidity. Performance of the VLEDs has been quantified using the measured luminous flux and color shift of the VLEDs subjected to both thermal and humidity stresses under a forward current bias of 350 mA. Results indicate that contamination can result in pre-mature luminous flux degradation and color shift in LEDs.

  14. Investigation of control rod worth and nuclear end of life of BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of {sup 10}B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% {sup 10}B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in {sup 10}B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming.

  15. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.

    1991-04-15

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR (boiling water reactor) in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed.

  16. High-fidelity multiphysics simulation of BWR assembly with coupled TORT-TD/CTF

    Energy Technology Data Exchange (ETDEWEB)

    Magedanz, J. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., Reber Building, Univ. Park, PA 16802 (United States); Perin, Y. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS mbH, Forschungsinstitute, D-85748 Garching (Germany); Avramova, M. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., Reber Building, Univ. Park, PA 16802 (United States); Pautz, A.; Puente-Espel, F.; Seubert, A.; Sureda, A.; Velkov, K.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS mbH, Forschungsinstitute, D-85748 Garching (Germany)

    2012-07-01

    This paper describes the application of the coupled codes TORT-TD and CTF to the pin-by-pin modeling of a BWR fuel assembly with thermal-hydraulic feedback. TORT-TD, developed at GRS, is a time-dependent three dimensional discrete ordinates code. CTF is the PSU's improved version of the subchannel code COBRA-TF, which uses a two-fluid, three-field model to represent two-phase flow with entrained droplets, and is commonly applied to evaluate LWR safety margins. The coupled codes system TORT-TD/CTF, already applied to several PWR cases involving MOX, was adapted from PWR to BWR applications. The purpose of the research described in this paper is to verify the coupling for modeling two-phase flow at the pin cell level. Using an ATRIUM-10 assembly, the system's steady-state capabilities were tested on two cases: one without control blade insertion and another with partially inserted blades. The influence of the neutron absorber on local axial and radial parameters is presented. The description of an inlet flow reduction transient is an example for the time-dependent capability of the coupled system. (authors)

  17. Advanced research workshop "South Caucasus : making the best use of external assistance for stability building and for co-operation with NATO"

    Index Scriptorium Estoniae

    2003-01-01

    Leedu Kaitseministeeriumi poolt 22.-23. septembrini 2003 Vilniuses korraldatud seminaril "South Caucasus: making the best use of external assistance for stability building and for co-operation with NATO" arutati, kas ja kuidas on Lõuna-Kaukaasia regioonis võimalik rakendada Balti riikide julgeolekualase koostöö kogemusi

  18. Field scale geomechanical modeling for prediction of fault stability during underground gas storage operations in a depleted gas field in the Netherlands

    NARCIS (Netherlands)

    Orlic, B.; Wassing, B.B.T.; Geel, C.R.

    2013-01-01

    A geomechanical modeling study was conducted to investigate stability of major faults during past gas production and future underground gas storage operations in a depleted gas field in the Netherlands. The field experienced induced seismicity during gas production, which was most likely caused by t

  19. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  20. Planning guidance for nuclear-power-plant decontamination. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munson, L.F.; Divine, J.R.; Martin, J.B.

    1983-06-01

    Direct and indirect costs of decontamination are considered in the benefit-cost analysis. A generic form of the benefit-cost ratio is evaluated in monetary and nonmonetary terms, and values of dollar per man-rem are cited. Federal and state agencies that may have jurisiction over various aspects of decontamination and waste disposal activities are identified. Methods of decontamination, their general effectiveness, and the advantages and disadvantages of each are outlined. Dilute or concentrated chemical solutions are usually used in-situ to dissolve the contamination layer and a thin layer of the underlying substrate. Electrochemical techniques are generally limited to components but show high decontamination effectiveness with uniform corrosion. Mechanical agents are particularly appropriate for certain out-of-system surfaces and disassembled parts. These processes are catagorized and specific concerns are discussed. The treatment, storage, and disposal or discharge or discharge of liquid, gaseous, and solid wastes generated during the decontamination process are discussed. Radioactive and other hazardous chemical wastes are considered. The monitoring, treatment, and control of radioactive and nonradioactive effluents, from both routine operations and possible accidents, are discussed. Protecting the health and safety of personnel onsite during decontamination is of prime importance and should be considered in each facet of the decontamination process. The radiation protection philosophy of reducing exposure to levels as low as reasonably achievable should be stressed. These issues are discussed.

  1. Analysis of a German BWR core with TRACE/PARCS using different cross section sets

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, C., E-mail: Christoph.Hartmann@kit.edu [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Westinghouse Electric Germany GmbH, Mannheim (Germany); Sanchez, V.H. [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Tietsch, W. [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2011-07-01

    'Full text:' Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools as well as appropriate cross sections (XS) libraries depending on the individual thermal hydraulic state parameters. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running and user-friendly lattice codes such as Casco and Helios. At research institutions and universities, alternative tools to the commercial ones with full access to the source code as well as moderate cost are urgently needed. The Scale system is being developed and improved for lattice physics calculations of real core loading of Light Water Reactors (LWR). It represents a promising alternative to the commercial lattice codes. At Karlsruhe Institute of Technology (Kit) a computational route based on Scale/Triton/Newt for BWR core loading is under development. The generated XS-data sets have to be transformed in PMAXS-format for use in the reactor dynamic code PARCS. This task is performed by the module GenPMAXS being developed and tested at the Michigan University. To verify the computational route, a BWR fuel assembly depletion problem was calculated by PARCS and compared to the CASMO results. Since the SCALE/TRITON XS-file does actually not contain all required neutronic data, FORTRAN routines have been developed to incorporate the missing data e.g. the yields of Iodine, Xenon and Promethium into the XS-data sets in the PMAXS-format. The comparison of the results obtained with PARCS (using the corrected PMAXS file) and CASMO for the depletion problem exhibited a good agreement. Consequently, this approach was followed for the generation of a complete XS-set for a real BWR core to be used in subsequent transient analysis. Then 3D neutronic and thermal hydraulic core model were elaborated for a TRACE/PARCS analysis. The thermal hydraulic model is based on the 3D VESSEL

  2. Problem Analyses in the Operation of Stability Control Equipment in Ningdong Power Plant%宁东电厂安稳装置运行问题分析

    Institute of Scientific and Technical Information of China (English)

    刘武旭

    2012-01-01

    This paper introduces the functions of the stability control equipment in Guohua Ningdong Power Plant.The possible problems in the operation of the stability control equipment in the 330kv substation wring mode are analyzed and the corresponding measures are proposed.%简要介绍了国华宁东电厂安稳装置的作用、功能。针对宁东电厂330KV变电站接线方式,对安稳装置运行中可能出现的问题进行了分析并提出了解决方案。

  3. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors; Actualizacion de los costos de combustible nuclear de la recarga de equilibrio de los reactores ABWR y EPR

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, R.F. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: rortega@fi-b.unam.mx

    2008-07-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U{sub 3O}8 in January, 2005 to a maximum of US$137.00 dollars by Ib U{sub 3}O{sub 8} by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U{sub 3}O{sub 8} such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  4. Ideal Stabilization

    CERN Document Server

    Nesterenko, Mikhail

    2009-01-01

    We define and explore the concept of ideal stabilization. The program is ideally stabilizing if its every state is legitimate. Ideal stabilization allows the specification designer to prescribe with arbitrary degree of precision not only the fault-free program behavior but also its recovery operation. Specifications may or may not mention all possible states. We identify approaches to designing ideal stabilization to both kinds of specifications. For the first kind, we state the necessary condition for an ideally stabilizing solution. On the basis of this condition we prove that there is no ideally stabilizing solution to the leader election problem. We illustrate the utility of the concept by providing examples of well-known programs and proving them ideally stabilizing. Specifically, we prove ideal stabilization of the conflict manager, the alternator, the propagation of information with feedback and the alternating bit protocol.

  5. Optimization of fuel cells for BWR based in Tabu modified search; Optimizacion de celdas de combustible para BWR basada en busqueda Tabu modificada

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Palomera P, M.A. [Facultad de Ingenieria, UNAM, Posgrado en Ingenieria en Computacion, Circuito exterior s/n, Ciudad Universitaria, Mexico, D.F. (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2004-07-01

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

  6. Fuel design with low peak of local power for BWR reactors with increased nominal power; Diseno de un combustible con bajo pico de potencia local para reactores BWR con potencia nominal aumentada

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia C, R.; Montes, J.L.; Hernandez, J.L.; Ortiz, J.J.; Castillo, A. [ININ, 52750 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2006-07-01

    The Federal Commission of Electricity recently announcement the beginning of the works related with the increase of the power to 120% of the original nominal one in the Boiling Water Reactors (BWR) of the Laguna Verde Central (CLV): In the National Institute of Nuclear Research (ININ) are carried out studies of the impact on the design of the recharge of derived fuel of this increase. One of the main effects of the power increase type that it is promoting, is the increment of the flow of generated vapor, what takes, to a bigger fraction of vacuum in the core presenting increased values of the maximum fraction to the limit, so much of the ratio of lineal heat generation (XFLPD) as of the ratio of critic power (MFLCPR). In the made studies, it is found that these fractions rise lineally with the increase of the nominal power. Considering that the reactors of the CLV at the moment operate to 105% of the original nominal power, it would imply an increment of the order of 13.35% in the XFLPD and in the MFLCPR operating to a nominal power of 120% of the original one. This would propitiate bigger problems to design appropriately the fuel cycle and the necessity, almost unavoidable, of to resort to a fuel assembly type more advanced for the recharges of the cores. As option, in the ININ the feasibility of continuing using the same type of it fuel assembles that one has come using recently in the CLV, the type GE12 is analyzed. To achieve it was outlined to diminish the peak factor of local power (LPPF) of the power cells that compose the fuel recharge in 13.35%. It was started of a fuel design previously used in the recharge of the unit 1 cycle 12 and it was re-design to use it in the recharge design of the cycle 13 of the unit 1, considering an increase to 120% of the original power and the same requirements of cycle extension. For the re-design of the fuel assembly cell it was used the PreDiCeldas computer program developed in the ININ. It was able to diminish the LPPF

  7. Cause-Effect relationship of the Laguna Verde BWR power instability by empirical mode decomposition; Relacion efecto-causa de la inestabilidad de potencia del BWR de Laguna Verde por descomposicion modal empirica

    Energy Technology Data Exchange (ETDEWEB)

    Blazquez, J.; Ruiz, J.; Castillo, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2008-07-01

    The signals coming from natural phenomena are in essence non lineal and not stationary. A recent development, well-known as Empirical Mode Decomposition (EMD) it presents a novel focus that allows to represent in adaptative form non stationary signals as a sum of components of half zero. These components denominated Intrinsic Mode Functions (IMF) they help to the analysis of the frequency composition of unidimensional signals. The use of the EMD followed by the Hilbert transform of the IMFs it allows to carry out an analysis in time-frequency of the non lineal and not stationary data. This technique is known as the Hilbert Huang Transform (HHT). In this work a power instability event occurred in January 24, 1995 in the unit I of the nuclear power station of Laguna Verde (Mexico), corresponding to a BWR/5 is analyzed. When a Nuclear Plant suffers a power instability event, it is required obligatorily to explain to the Regulator Organism the effects and the causes of the event. The effects are described simply; not in vain there is a registration of signals in the Process Computer of where the required information is extracted. But the causes are not always immediate and easy for to identify. The power instability can happen during the start, when the refrigeration flow is relatively low in front to the power. By reason of that the reactivity coefficient by holes is negative, the power oscillates with a very defined frequency, generally of the order of 0.5 Hz. If the oscillations increase progressively of amplitude, we are in an instability event. It is interesting to include in the report the instant in that the began instability and the actions of the operator before and after the same one. As the actions are registered, the investigation is focused toward the instant of the beginning to be able to identify them. In this work the power signal in five empiric ways of Hilbert-Huang and a residual breaks down. The instability is only reflected in the way of smaller

  8. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  9. Study of intermediate configurations during the fuel reload in BWRs; Estudio de configuraciones intermedias durante la recarga de combustible en BWR's

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia del C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Jacinto C, S., E-mail: luis.fuentes@inin.gob.mx [Universidad Autonoma del Estado de Yucatan, Calle 60 No. 491-A por 57, 97000 Merida, Yucatan (Mexico)

    2012-10-15

    The criticality state of the core of a boiling water reactor (BWR) was evaluated, during the reload process for the intermediate states between the load pattern of cycle end and the beginning of the next, using the information of the load pattern of the operation cycles 13 and 14 of Unit 1 of the nuclear power plant of Laguna Verde. For this evaluation the codes CASMO-4 and Simulate-3 for conditions of the core in cold were used. The strategy consisted on moving assemblies with 4 burned cycles of the reactor core. Later on were re situated the remaining assemblies, placing them in the positions to occupy in the next operation cycle. Finally, was carried out the assemblies load of fresh fuel. In each realized change, it was observing the behavior of the k-effective value that is the parameter used to evaluate the criticality state of each state of the core change. In a second stage, was designed a program that builds in automatic way each one of the intermediate cores and also analyzes the criticality state of the reactor core after each withdrawal, re situated and load of fuel assemblies. (Author)

  10. Non-local two phase flow momentum transport in S BWR

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Salinas M, L.; Vazquez R, A., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Apdo. Postal 55-535, 09340 Ciudad de Mexico (Mexico)

    2015-09-15

    The non-local momentum transport equations derived in this work contain new terms related with non-local transport effects due to accumulation, convection, diffusion and transport properties for two-phase flow. For instance, they can be applied in the boundary between a two-phase flow and a solid phase, or in the boundary of the transition region of two-phase flows where the local volume averaging equations fail. The S BWR was considered to study the non-local effects on the two-phase flow thermal-hydraulic core performance in steady-state, and the results were compared with the classical local averaging volume conservation equations. (Author)

  11. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  12. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  13. Kuosheng BWR/6 containment pressure and temperature responses after recirculation line break using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lin, A.; Wang, J-R.; Chen, Y-S., E-mail: samuellin1999@iner.gov.tw, E-mail: jrwang@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research Atomic Energy Council (China); Shih, C., E-mail: ckshih@ess.nthu.edu.tw [National Tsing Hua Univ., Dept. of Engineering and System Science (China)

    2011-07-01

    In this study, we presented the calculated results of the containment P/T (pressure and temperature) response after the recirculation line break (RCLB) accident of a GE-designed twin-unit BWR/6 plant, which can be served as the design basis for the containment system. During the simulation, a power of SPU (stretch power uprate) range was used and a model of the Mark III type containment was built using the GOTHIC (Generation of Thermal-Hydraulic Information for Containments) code. The calculated results, similar to the FSAR (Final Safety Analysis Report) results, indicate the GOTHIC code has the capability to simulate the containment P/T response to the RCLB accident. (author)

  14. CSAU methodology and results for an ATWS event in a BWR using information theory methods

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J.L., E-mail: jlcobos@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Escrivá, A., E-mail: aescriva@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Mendizabal, R., E-mail: rmsanz@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Pelayo, F., E-mail: fpl@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Melara, J., E-mail: jls@iberdrola.es [IBERINCO, IBERDROLA Ingeniería y Construcción, Madrid (Spain)

    2014-10-15

    Highlights: • We apply the CSAU methodology to an ATWS in a BWR using information theory methods. • We show how to perform the selection of the most influential inputs on the critical safety parameter. • We apply the maximum entropy principle to get the input parameter distribution. • We examine the maximum relative entropy principle to update the input parameter PDF. • We quantify the uncertainty of the critical safety parameter using order statistics and information theory. - Abstract: This paper shows an application of the CSAU methodology to an ATWS in a BWR reactor, when the temperature of the suppression pool is taken as the critical safety parameter. The method combines CSAU methodology with recent techniques of information theory. In this paper we use auxiliary tools to help in the evaluation and improvement of the parameters distribution that enter in the elements II and III of CSAU based methodologies. These tools have been implemented in two FORTRAN programs: GEDIPA (Generation of the Parameter Distribution) and UNTHERCO (Uncertainty in Thermal Hydraulic Codes). The first one analyzes the information data available on a given parameter or parameters with the goal to know all the information about the probability distribution function of these parameters. The second apply information theory methods, as the maximum entropy principle (MEP) and the maximum relative entropy Principle (MREP), in order to build conservative distribution functions for the parameters from the available data. Also, the distribution function of a given parameter can be updated using the MREP principle when new information is provided. UNTHERCO performs the MONTECARLO sampling for a given set of parameters when the distribution function of these parameters is previously known. If the distribution of a parameter is unknown, then, the MEP is applied to deduce the distribution function for this parameter.

  15. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Chopra, O. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Gruber, Eugene E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shack, William J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  16. Influence of iron and nickel species upon activity buildup under simulated BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bjornsson, S.; Chen, J. [Studsvik Nuclear AB, Nykoping (Sweden); Lejon, J. [OKG AB, Oskarshamn (Sweden); Granath, G. [Ringhals AB, Varobacka (Sweden); Tanse-Larsson, M. [Forsmarks Kraftgrupp AB, Osthammar (Sweden)

    2010-07-01

    Activity build-up in BWR systems are of importance for service- and maintenance work performed at the plants. Minimizing the activity build-up is desirable for minimizing doses of personnel at the plants. Numerous studies have been carried out in this important field to understand the activity uptake mechanisms. This paper studied the possible role of Fe(II/III) and Ni(II) impurities in reactor water in activity uptake on stainless steel surfaces. The study was carried out by using a test loop with simulated BWR water containing Fe(II/III), Ni(II) and Co-60 marked Co(II) species of varied concentration and 500 ppb O{sub 2}. The test tube section in the loop system was pre-exposed type 316L stainless steel material. The microstructures of the formed oxide films were examined with high resolution electron microscopy (FE-SEM and FE-TEM). The activity monitoring on the test section showed that injection of 10 ppb Ni(II) and 0.1 ppb Fe(II/III) in the water with 0.1 ppb Co(II) was capable of stopping completely activity uptake. When Co(II) addition in the loop was stopped no activity return to the water could be seen. In another exposure test, injection of combined 2 ppb Fe(II/III) and 0.5∼10 ppb Ni(II) profoundly increased activity uptake on the test section with a maximum in activity buildup at 5 ppb Ni(II). When Co(II) addition in the loop was stopped a slight activity return was seen. The observed differences as seen in the two tests are discussed in view of the microstructures of the oxide films formed. (author)

  17. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  18. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  19. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  20. Analysis of the performance of fuel cells BWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles BWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Vargas, S.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: govaj666@hotmail.com

    2008-07-01

    The efficient use of the fuel is one of the objectives in the assemblies design of type BWR. The present tendency in the assemblies design of type BWR is through a radial distribution of enrichments. The present work has like object showing the because of this decision, for what a comparison of the neutronic performance of two fuel cells with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The cells were analyzed with the CASMO-4 code and the obtained results of the behavior of the neutron flow and the power sustain the because of the radial distribution of enrichments. (Author)

  1. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type; Calculo de las razones de generacion de calor lineal que violen el limite termomecanico de deformacion plastica de la camisa en funcion del quemado de una barra combustible tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2003-07-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  2. [The pathogenetic approach to the development of tools and methods for the improvement of statokinetic stability in the operators of aerospace systems].

    Science.gov (United States)

    Glaznikov, L A; Buĭnov, L G; Govorun, M I; Sorokina, L A; Nigmedzianov, R A; Golovanov, A E

    2012-01-01

    The objective of the present study was to estimate the efficacy of the tools and methods for the optimization of the activity of the central nervous system (CNS) and analyzers involved in the maintenance of the statokinetic (SK) stability in man. To this effect, we evaluated the outcome of bemitil treatment during 10 days with and without A.I. Yarotsky test and the influence of these procedures on the pathophysiological characteristics of selected elements of the work of operators of aerospace systems. Based on the data obtained in the study, the tools and methods have been developed that allow the efficacy and quality of certain aspects of the operators' activity to be improved, viz. general working capacity under conditions requiring enhanced statokinetic stability, self-confidence, emotional and somatic comfort.

  3. The general as statesman : exploring the professional need for commanders to support viable political outcomes in peace and stability operations as typified by the UK military approach

    OpenAIRE

    Lovelock, R

    2011-01-01

    The problem of theatre level politico-military arrangements during peace and stability operations is important because the intervening actors, working in complex and often ambiguous circumstances, need to calibrate the application of military and political means as a coherent interdependent whole. This is necessary in order to build peace, secure viable political outcomes and hence strategic successes; however it is not easy in practice. This thesis examines the hypothesis that, beyond their ...

  4. Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR; Foerstudie av stroemningsinducerade laster paa interndelar vid brott i huvudcirkulationskretsarna i BWR

    Energy Technology Data Exchange (ETDEWEB)

    Marcinkiewicz, Jerzy; Lindgren, Anders [Det Norske Veritas Nuclear Technology AB, Stockholm (Sweden)

    2002-12-01

    Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report.

  5. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR; Calculo de flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.

    2011-07-01

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

  6. Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Urquidi-Macdonald, Mirna [Penn State University, 212 Earth-Engineering Science Building, University Park, PA 16801 (United States)

    2004-07-01

    Many publications have shown that crack growth rates (CGR) due to intergranular stress corrosion cracking (IGSCC) of metals is dependent on many parameters related to the manufacturing process of the steel and the environment to which the steel is exposed. Those parameters include, but are not restricted to, the concentration of chloride, fluoride, nitrates, and sulfates, pH, fluid velocity, electrochemical potential (ECP), electrolyte conductivity, stress and sensitization applied to the steel during its production and use. It is not well established how combinations of each of these parameters impact the CGR. Many different models and beliefs have been published, resulting in predictions that sometimes disagree with experimental observations. To some extent, the models are the closest to the nature of IGSCC, however, there is not a model that fully describes the entire range of observations, due to the difficulty of the problem. Among the models, the Fracture Environment Model, developed by Macdonald et al., is the most physico-chemical model, accounting for experimental observations in a wide range of environments or ECPs. In this work, we collected experimental data on BWR environments and designed a data mining pattern recognition model to learn from that data. The model was used to generate CGR estimations as a function of ECP on a BWR environment. The results of the predictive model were compared to the Fracture Environment Model predictions. The results from those two models are very close to the experimental observations of the area corresponding to creep and IGSCC controlled by diffusion. At more negative ECPs than the potential corresponding to creep, the pattern recognition predicts an increase of CGR with decreasing ECP, while the Fracture Environment Model predicts the opposite. The results of this comparison confirm that the pattern recognition model covers 3 phenomena: hydrogen embrittlement at very negative ECP, creep at intermediate ECP, and IGSCC

  7. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  8. Possibilities with OHWC. Development and application of ECP-simulation in Swedish BWRs; Moejligheter med OHWC. Utveckling och tillaempning av ECP-simulering i svenska BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lundgren, K. [ALARA Engineering, Skultuna (Sweden); Wikmark, G. [Advanced Nuclear Technology, Uppsala (Sweden)

    2000-02-01

    Hydrogen injection (HWC) to boiling water reactors has been used for two decades in Sweden, in order to reduce the impact of pipe cracking. The effect of HWC is to establish a sufficiently reducing environment in the systems to protect and hence mitigate the growth of existing stress corrosion cracks. Some disadvantages of HWC have been identified. One is the transitional increase of the dose rate of the main steam lines by up to seven times, another the corrosion release of systems with carbon steel components as a result of the reducing chemistry. In some cases, especially in the USA, an elevated activity build-up has been observed in a few plants in connection to the application of HWC. There is also a fear for increased hydrogen pick-up in fuel cladding and fuel channels by HWC operation. The hydrogen pick-up is already today in many cases limiting for fuel life. The objective of the current work has been to investigate the conditions by application of so called Optimised HWC. This implies a HWC operation with lower hydrogen addition rates than normally used. For this purpose, a computer model in order to simulate the radiolysis chemistry and the ECP (electrochemical corrosion potentials) in BWR systems has been developed. A previously developed radiolysis code, BwrChem, as well as a hydrogen peroxide decomposition code for piping, PEROX, have hence been equipped with ECP calculation modules. The ECP calculation algorithms have been based on fundamental electrochemical theory. The new model has been applied to simulate the radiolysis conditions in a large number of locations in typical BWRs. For the simulation, the external mechanical pump plant Barsebaeck-1 and the internal pump plant Forsmark-1 have been used. A wide range of hydrogen injection rates, down to 0. 1 ppm in the feed water, have been studied. The electrochemical model based on fundamental theory required adequate fundamental parameters. Significant effort has been used to scrutinise and evaluate

  9. Development of a fuzzy logic method to build objective functions in optimization problems: application to BWR fuel lattice design

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, C.; Francois, J.L.; Barragan, A.M. [Universidad Nacional Autonoma de Mexico - Facultad de Ingenieria (Mexico); Palomera, M.A. [Universidad Nacional Autonoma de Mexico - Instituto de Investigaciones en Matematicas Aplicadas y Sistema, Mexico, D. F. (Mexico)

    2005-07-01

    In this paper we develop a methodology based on the use of the Fuzzy Logic technique to build multi-objective functions to be used in optimization processes applied to in-core nuclear fuel management. As an example, we selected the problem of determining optimal radial fuel enrichment and gadolinia distributions in a typical 'Boiling Water Reactor (BWR)' fuel lattice. The methodology is based on the use of the mathematical capability of Fuzzy Logic to model nonlinear functions of arbitrary complexity. The utility of Fuzzy Logic is to map an input space into an output space, and the primary mechanism for doing this is a list of if-then statements called rules. The rules refer to variables and adjectives that describe those variables and, the Fuzzy Logic technique interprets the values in the input vectors and, based on the set of rules assigns values to the output vector. The methodology was developed for the radial optimization of a BWR lattice where the optimization algorithm employed is Tabu Search. The global objective is to find the optimal distribution of enrichments and burnable poison concentrations in a 10*10 BWR lattice. In order to do that, a fuzzy control inference system was developed using the Fuzzy Logic Toolbox of Matlab and it has been linked to the Tabu Search optimization process. Results show that Tabu Search combined with Fuzzy Logic performs very well, obtaining lattices with optimal fuel utilization. (authors)

  10. Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. del C.

    2015-07-01

    A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

  11. Mine operating accurate stability control with optical fiber sensing and Bragg grating technology: the BRITE-EURAM STABILOS project

    Science.gov (United States)

    Ferdinand, Pierre; Ferragu, Olivier; Lechien, J. L.; Lescop, B.; Marty-DeWinter, Veronique; Rougeault, S.; Pierre, Guillaume; Renouf, C.; Jarret, Bertrand; Kotrotsios, Georges; Neuman, Victor; Depeursinge, Y.; Michel, J. B.; Van Uffelen, M.; Verbandt, Yves; Voet, Marc R. H.; Toscano, D.

    1994-09-01

    Recent developments of stability control in mines, essentially based on Ge-doped Fiber Bragg Gratings (FBG) are reported including results about the different aspects of the system: accurate characterizations of FBG, sensor network topology and multiplexing method, user interface design and sensor packaging.

  12. Monitoring the NOAA Operational VIIRS RSB and DNB Calibration Stability Using Monthly and Semi-Monthly Deep Convective Clouds Time Series

    Directory of Open Access Journals (Sweden)

    Wenhui Wang

    2016-01-01

    Full Text Available The Visible and Infrared Imaging Radiometer Suite (VIIRS onboard the Joint Polar Satellite System (JPSS/Suomi National Polar-Orbiting Partnership (SNPP satellite provide sensor data records for the retrievals of many environment data records. It is critical to monitor the VIIRS long-term calibration stability to ensure quality EDR retrieval. This study investigates the radiometric calibration stability of the NOAA operational SNPP VIIRS Reflective Solar Bands (RSB and Day-Night-Band (DNB using Deep Convective Clouds (DCC. Monthly and semi-monthly DCC time series for 10 moderate resolution bands (M-bands, M1–M5 and M7–M11, March 2013–September 2015, DNB (March 2013–September 2015, low gain stage, and three imagery resolution bands (I-bands, I1–I3, January 2014–September 2015 were developed and analyzed for long-term radiometric calibration stability monitoring. Monthly DCC time series show that M5 and M7 are generally stable, with a stability of 0.4%. DNB has also been stable since May 2013, after its relative response function update, with a stability of 0.5%. The stabilities of M1–M4 are 0.6%–0.8%. Large fluctuations in M1–M4 DCC reflectance were observed since early 2014, correlated with F-factor (calibration coefficients trend changes during the same period. The stabilities of M8-M11 are from 1.0% to 3.1%, comparable to the natural DCC variability at the shortwave infrared spectrum. DCC mean band ratio time series show that the calibration stabilities of I1–I3 follow closely with M5, M7, and M10. Relative calibration changes were observed in M1/M4 and M5/M7 DCC mean band ratio time series. The DCC time series are generally consistent with results from the VIIRS validation sites and VIIRS/MODIS (the Moderate-resolution Imaging Spectroradiometer simultaneous nadir overpass time series. Semi-monthly DCC time series for RSB M-bands and DNB were compared with monthly DCC time series. The results indicate that semi-monthly DCC

  13. A robust method for frequency stabilization of 556-nm laser operating at the intercombination transition of ytterbium

    Institute of Scientific and Technical Information of China (English)

    Zhuanxian Xiong; Yun Long; Huixing Xiao; Xi Zhang; Lingxiang He; Baolong Lu

    2011-01-01

    A frequency-stabilized 556-nm laser is an essential tool for experimental studies associated with 1S0-3p1 intercombination transition of ytterbium (Yb) atoms. A 556-nm laser light using a single-pass second harmonic generation (SHG) is obtained in a periodically poled MgO:LiNbO3 (PPLN) crystal pumped by a fiber laser at 1111.6 nm. A robust frequency stabilization method which facilitates the control of laser frequency with an accuracy better than the natural linewidth (187 kHz) of the intercombination line is developed. The short-term frequency jitter is reduced to less than 100 kHz by locking the laser to a home-made reference cavity. A slow frequency drift is sensed by the 556-nm fluorescence signal of an Yb atomic beam excited by one probe beam and is reduced to less than 50-kHz by a computer-controlled servo system. The laser can be stably locked for more than 5 h. This frequency stabilization method can be extended to other alkaline-earth-like atoms with similar weak intercombination lines.%@@ A frequency-stabilized 556-nm laser is an essential tool for experimental studies associated with 1So-3P1intercombination transition of ytterbium (Yb) atoms.A 556-nm laser light using a single-pass second harmonic generation (SHG) is obtained in a periodically poled MgO:LiNbO3 (PPLN) crystal pumped by a fiber laser at 1111.6 nm.A robust frequency stabilization method which facilitates the control of laser frequency with an accuracy better than the natural linewidth (187 kHz) of the intercombination line is developed.The short-term frequency jitter is reduced to less than i00 kHz by locking the laser to a home-made reference cavity.A slow frequency drift is sensed by the 556-nm fluorescence signal of an Yb atomic beam excited by one probe beam and is reduced to less than 50-kHz by a computer-controlled servo system.The laser can be stably locked for more than 5 h.This frequency stabilization method can be extended to other alkaline-earth-like atoms with similar weak

  14. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  15. A post-processing technique for stabilizing the discontinuous pressure projection operator in marginally-resolved incompressible inviscid flow

    CERN Document Server

    Joshi, Sumedh M; Steinmoeller, Derek T; Stastna, Marek; Thomsen, Greg N

    2015-01-01

    A method for post-processing the velocity after a pressure projection is developed that helps to maintain stability in an under-resolved, inviscid, discontinuous element-based simulation for use in environmental fluid mechanics process studies. The post-processing method is needed because of spurious divergence growth at element interfaces due to the discontinuous nature of the discretization used. This spurious divergence eventually leads to a numerical instability. Previous work has shown that a discontinuous element-local projection onto the space of divergence-free basis functions is capable of stabilizing the projection method, but the discontinuity inherent in this technique may lead to instability in under-resolved simulations. By enforcing inter-element discontinuity and requiring a divergence-free result in the weak sense only, a new post-processing technique is developed that simultaneously improves smoothness and reduces divergence in the pressure-projected velocity field at the same time. When com...

  16. Assessment of the performance of Ni-yttria-stabilized zirconia anodes in anode-supported Solid Oxide Fuel Cells operating on H 2-CO syngas fuels

    Science.gov (United States)

    Ye, Xiao-Feng; Wang, S. R.; Zhou, J.; Zeng, F. R.; Nie, H. W.; Wen, T. L.

    Anode-supported Solid Oxide Fuel Cells (SOFCs) with Ni-yttria-stabilized zirconia (YSZ) anode have been fabricated and studied using H 2-CO syngas fuels. Syngas fuels with different compositions of H 2-CO are supplied and the cell performance is measured at 750 °C. A high CO content has caused carbon deposition and crack formation in the Ni-YSZ anode after long-term operation, even though it is diluted with H 2O and N 2. However, it was found that a Cu-CeO 2 coating on Ni-YSZ can greatly improve the anode stability in syngas by facilitating the water gas shift reaction. The optimized single cell has run in sygas with a composition of 65%H 2-32%CO-3%H 2O for 1050 h without obvious degradation of its performance.

  17. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

  18. Sharp three sphere inequality for perturbations of a product of two second order elliptic operators and stability for the Cauchy problem for the anisotropic plate equation

    CERN Document Server

    Morassi, Antonino; Vessella, Sergio

    2010-01-01

    We prove a sharp three sphere inequality for solutions to third order perturbations of a product of two second order elliptic operators with real coefficients. Then we derive various kinds of quantitative estimates of unique continuation for the anisotropic plate equation. Among these, we prove a stability estimate for the Cauchy problem for such an equation and we illustrate some applications to the size estimates of an unknown inclusion made of different material that might be present in the plate. The paper is self-contained and the Carleman estimate, from which the sharp three sphere inequality is derived, is proved in an elementary and direct way based on standard integration by parts.

  19. Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

    Science.gov (United States)

    Huang, J. Y.; Chiang, M. F.; Jeng, S. L.; Huang, J. S.; Kuo, R. C.

    2013-01-01

    The environmentally assisted cracking behavior of dissimilar metal (DM) welds, including Alloy 52-A 508 and Alloy 82-A508, under simulated BWR coolant conditions was studied. Effects of postweld heat treatment and sulfur content of the base metal on the corrosion fatigue and SCC growth rates of DM welds were evaluated. The crack growth rates for the DM weld heat-treated at 621 °C for 24 h were observed to be faster than those for the as-welded. But the DM weld heat-treated at 621 °C for 8 h + 400 °C for 200 h showed better SCC resistance than the as-welded. The longer the heat treatment at 621 °C, the higher the chromium carbides density along the grain boundary was observed. Sulfur could diffuse out of the base metal and segregate along the grain boundaries of the dilution zone, leading to weakening the grain boundary strength and the SCC resistance of the Alloy 52-A508 weld.

  20. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  1. Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies

    Directory of Open Access Journals (Sweden)

    Andrius Slavickas

    2015-01-01

    Full Text Available The study of the void reactivity variation in innovative BWR fuel assemblies is presented in this paper. The innovative assemblies are loaded with high enrichment fresh UO2 and MOX fuels. UO2 fuel enrichment is increased above existing design limitations for LWR fuels (>5%. MOX fuel enrichment with fissile Pu content is established to achieve the same burnup level as that of high enrichment UO2 fuel. For the numerical analysis, the TRITON functional module of SCALE 6.1 code with the 238-group ENDF/B-VI cross section data library was applied. The investigation of the void reactivity feedback is performed in the entire 0–100% void fraction range. Higher values of void reactivity coefficient for assembly loaded with MOX fuel are found in comparison with values for assembly loaded with UO2 fuel. Moreover, coefficient values for MOX fuel are positive over 75% void fraction. The variation of the void reactivity coefficient is explained by the results of the decomposition analysis based on four-factor formula and neutron absorption reactions for main isotopes. Additionally, the impact of the moderation enhancement on the void reactivity coefficient was investigated for the innovative assembly with MOX fuel.

  2. Decomposition Analysis of Void Reactivity Coefficient for Innovative and Modified BWR Assemblies

    Directory of Open Access Journals (Sweden)

    Andrius Slavickas

    2014-01-01

    Full Text Available The decomposition analysis of void reactivity coefficient for innovative BWR assemblies is presented in this paper. The innovative assemblies were loaded with high enrichment UO2 and MOX fuels. Additionally the impact of the moderation enhancement on the void reactivity coefficient through a full fuel burnup discharge interval was investigated for the innovative assembly with MOX fuel. For the numerical analysis the TRITON functional module of SCALE code with ENDF/B-VI cross section library was applied. The obtained results indicate the influence of the most important isotopes to the void reactivity behaviour over a fuel burnup interval of 70 GWd/t for both UO2 and MOX fuels. From the neutronic safety concern positive void reactivity coefficient values are observed for MOX fuel at the beginning of the fuel irradiation cycle. For extra-moderated assembly designs, implementing 8 and 12 water holes, the neutron spectrum softening is achieved and consequently the lower void reactivity values. Variations in void reactivity coefficient values are explained by fulfilled decomposition analysis based on neutrons absorption reactions for separate isotopes.

  3. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

    1999-07-01

    This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

  4. Thermal utilization opportunities with a small-to-medium sized BWR

    Energy Technology Data Exchange (ETDEWEB)

    Konkin, D.; Simonson, C.J.; Dalai, A.K.; Tanino, K.; Guo, H., E-mail: doug.konkin@usask.ca [Univ. of Saskatchewan, Saskatchewan (Canada); Nishida, K.; Mochida, T. [Hitachi-GE Nuclear Energy, Ltd., Ibaraki (Japan); Ikegawa, T.; Kito, K. [Hitachi, Ltd., Ibaraki (Japan); Knudsen, R. [LeanOptions Consulting, Inc., Regina, Saskatchewan (Canada); Aikman, A. [SNC-Lavalin, Saskatoon, Saskatchewan (Canada); Humphries, R. [AMEC, Toronto, Ontario (Canada)

    2014-07-01

    Hitachi-GE Nuclear Energy Ltd. (Hitachi-GE) has developed a conceptual design for a Double MS: Modular Simplified & Medium Small Reactor (DMS) under the sponsorship of The Japan Atomic Power Company. Recent efforts have yielded enhancements for improved safety and reactor core performance. The DMS is an innovative small-to-medium sized Boiling Water Reactor (BWR), which, based only on electricity generation, has been estimated to almost overcome economy of scale concerns when compared to proven conventional Advanced Boiling Water Reactor (ABWR) technologies. In order to make the DMS more attractive, the University of Saskatchewan (U of S), Hitachi-GE and Hitachi Ltd. (Hitachi) have collaborated on a joint research and development (R&D) initiative to study the utilization of heat and steam from the Balance of Plant (BOP) associated with the DMS for thermal utilization (TU) applications. In this paper, the advanced features of the DMS and the individual projects of the R&D program will be described. (author)

  5. Mode shift and stability control of a current mode controlled buck-boost converter operating in discontinuous conduction mode with ramp compensation

    Institute of Scientific and Technical Information of China (English)

    Bao Bo-Cheng; Xu Jian-Ping; Liu Zhong

    2009-01-01

    By establishing the discrete iterative mapping model of a current mode controlled buck-boost converter, this paper studies the mechanism of mode shift and stability control of the buck-boost converter operating in discontinuous conduction mode with a ramp compensation current. With the bifurcation diagram, Lyapunov exponent spectrum, timedomain waveform and parameter space map, the performance of the buck-boost converter circuit utilizing a compensating ramp current has been analysed. The obtained results indicate that the system trajectory is weakly chaotic and strongly intermittent under discontinuous conduction mode. By using ramp compensation, the buck-boost converter can shift from discontinuous conduction mode to continuous conduction mode, and effectively operates in the stable period-one region.

  6. GENERAL: Mode shift and stability control of a current mode controlled buck-boost converter operating in discontinuous conduction mode with ramp compensation

    Science.gov (United States)

    Bao, Bo-Cheng; Xu, Jian-Ping; Liu, Zhong

    2009-11-01

    By establishing the discrete iterative mapping model of a current mode controlled buck-boost converter, this paper studies the mechanism of mode shift and stability control of the buck-boost converter operating in discontinuous conduction mode with a ramp compensation current. With the bifurcation diagram, Lyapunov exponent spectrum, time-domain waveform and parameter space map, the performance of the buck-boost converter circuit utilizing a compensating ramp current has been analysed. The obtained results indicate that the system trajectory is weakly chaotic and strongly intermittent under discontinuous conduction mode. By using ramp compensation, the buck-boost converter can shift from discontinuous conduction mode to continuous conduction mode, and effectively operates in the stable period-one region.

  7. Ionic Liquid-Organic Carbonate Electrolyte Blends To Stabilize Silicon Electrodes for Extending Lithium Ion Battery Operability to 100 °C.

    Science.gov (United States)

    Ababtain, Khalid; Babu, Ganguli; Lin, Xinrong; Rodrigues, Marco-Tulio F; Gullapalli, Hemtej; Ajayan, Pulickel M; Grinstaff, Mark W; Arava, Leela Mohana Reddy

    2016-06-22

    Fabrication of lithium-ion batteries that operate from room temperature to elevated temperatures entails development and subsequent identification of electrolytes and electrodes. Room temperature ionic liquids (RTILs) can address the thermal stability issues, but their poor ionic conductivity at room temperature and compatibility with traditional graphite anodes limit their practical application. To address these challenges, we evaluated novel high energy density three-dimensional nano-silicon electrodes paired with 1-methyl-1-propylpiperidinium bis(trifluoromethanesulfonyl)imide (Pip) ionic liquid/propylene carbonate (PC)/LiTFSI electrolytes. We observed that addition of PC had no detrimental effects on the thermal stability and flammability of the reported electrolytes, while largely improving the transport properties at lower temperatures. Detailed investigation of the electrochemical properties of silicon half-cells as a function of PC content, temperature, and current rates reveal that capacity increases with PC content and temperature and decreases with increased current rates. For example, addition of 20% PC led to a drastic improvement in capacity as observed for the Si electrodes at 25 °C, with stability over 100 charge/discharge cycles. At 100 °C, the capacity further increases by 3-4 times to 0.52 mA h cm(-2) (2230 mA h g(-1)) with minimal loss during cycling.

  8. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

    2012-07-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

  9. DOD Task Force for Business and Stability Operations in Afghanistan: Review of Selected Expenditures Highlights Serious Management and Oversight Problems

    Science.gov (United States)

    2016-04-15

    Brinkley in December 2015, the provisions of the fiscal year 2011 National Defense SIGAR 16-29-TY Page 7 Mr. Brinkley also told SIGAR that without...pipeline, to be liabilities due to safety concerns, lack of sustainability, and other problems.38 TFBSO’s last director, Dr. Joseph Catalino, confirmed...Congress and future administrations when considering economic development activities in future contingency operations. To date, SIGAR has not been

  10. U.S. Border Patrol OCONUS: Possible Contributions to the Whole of Government Approach to Stability Operations

    Science.gov (United States)

    2010-06-11

    Agents (ACPA) Randall Baldwin, Gerald S. Bryan, and Anthony Porvaznik; Patrol Agents in Charge Gregory Bovino , Desi D. DeLeon, Vincent J. Hampel, and...leaders to improve relations with the DBE. 94 In 2001, Gregory Bovino was a Supervisory Border Patrol Agent (SBPA) with the BORTAC operating in...Honduras. Gregory Bovino is now a PAIC. PAIC Bovino helped organize and train host-nation border security forces in Honduras from the ground up through

  11. Optimization of analysis best-estimate of a fuel element BWR with Code STAR-CCM+; Optimizacion del analisis best-estimate de un elemento combustible BWR con el codigo STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Morgado Canada, E.; Concejal Barmejo, A.; Jimenez Varas, G.; Solar Martinez, A.

    2014-07-01

    The objective of the project is the evaluation of the code STAR-CCM +, as well as the establishment of guidelines and standardized procedures for the discretization of the area of study and the selection of physical models suitable for the simulation of BWR fuel. For this purpose several of BFBT experiments have simulated [1] provide a data base for the development of experiments for measuring distribution of fractions of holes to changes in power in order to find the most appropriate models for the simulation of the problem. (Author)

  12. Long-term stability of underground operated CZT detectors based on the analysis of intrinsic {sup 113}Cd β{sup −}-decay

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, J. [Universität Hamburg, Institut für Experimentalphysik, Luruper Chaussee 149, 22761 Hamburg (Germany); Gößling, C. [Technische Universität Dortmund, Lehrstuhl für Experimentelle Physik IV, Otto-Hahn-Str 4, 44221 Dortmund (Germany); Gehre, D., E-mail: daniel.gehre@tu-dresden.de [Technische Universität Dresden, Institut für Kern- und Teilchenphysik, Zellescher Weg 19, 01069 Dresden (Germany); Hagner, C.; Heidrich, N. [Universität Hamburg, Institut für Experimentalphysik, Luruper Chaussee 149, 22761 Hamburg (Germany); Klingenberg, R.; Kröninger, K.; Nitsch, C. [Technische Universität Dortmund, Lehrstuhl für Experimentelle Physik IV, Otto-Hahn-Str 4, 44221 Dortmund (Germany); Oldorf, C. [Universität Hamburg, Institut für Experimentalphysik, Luruper Chaussee 149, 22761 Hamburg (Germany); Quante, T.; Rajek, S. [Technische Universität Dortmund, Lehrstuhl für Experimentelle Physik IV, Otto-Hahn-Str 4, 44221 Dortmund (Germany); Rebber, H.; Rohatsch, K. [Universität Hamburg, Institut für Experimentalphysik, Luruper Chaussee 149, 22761 Hamburg (Germany); Tebrügge, J.; Temminghoff, R.; Theinert, R. [Technische Universität Dortmund, Lehrstuhl für Experimentelle Physik IV, Otto-Hahn-Str 4, 44221 Dortmund (Germany); and others

    2016-06-11

    The COBRA collaboration operates a demonstrator setup at the underground facility Laboratori Nazionali del Gran Sasso (LNGS, located in Italy) to prove the technological capabilities of this concept for the search for neutrinoless double beta-decay. The setup consists of 64 (1×1×1) cm{sup 3} Cadmium-Zinc-Telluride (CZT) detectors in Coplanar-Grid (CPG) configuration. One purpose of this demonstrator is to test if reliable long-term operation of CZT-CPG detectors in such a setup is possible. The demonstrator has been operated under ultra low-background conditions for more than three years and collected data corresponding to a total exposure of 218 kg days. The presented study focuses on the long-term stability of CZT detectors by analyzing the intrinsic, fourfold forbidden non-unique {sup 113}Cd single beta-decay. It can be shown that CZT detectors can be operated stably for long periods of time and that the {sup 113}Cd single beta-decay can be used as an internal monitor of the detector performance during the runtime of the experiment.

  13. Effect of organic load and nutrient ratio on the operation stability of the moving bed bioreactor for kraft mill wastewater treatment and the incidence of polyhydroxyalkanoate biosynthesis.

    Science.gov (United States)

    Pozo, G; Villamar, C A; Martínez, M; Vidal, G

    2012-01-01

    This paper studies the effect of organic load rate (OLR) and nutrient ratio on operation stability of the moving bed bioreactor (MBBR) for kraft mill wastewater treatment, analyzing the incidence of polyhydroxyalkanoate (PHA) production. The MBBR operating strategy was to increase OLR from 0.25 ± 0.05 to 2.41 ± 0.19 kg COD m(-3) d(-1) between phases I and IV. The BOD(5):N:P ratio (100:5:1 and 100:1:0.2) was evaluated as an operation strategy for phases IV to V. A stable MBBR operation was found when the OLR was increased during 225 days in five phases. The maximum absolute fluorescence against the proportion of cells accumulating PHA was obtained for an OLR of 2.41 ± 0.19 kg COD m(-3)d(-1) and a BOD(5):N:P relationship of 100:1:0.2. The increase of PHA biosynthesis is due to the increased OLR and is not attributable to the increased cell concentration, which is maintained constant in stationary status during bioreactor biosynthesis.

  14. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

    Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  15. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133{degrees}F) has a 95-percentile uncertainty of 14.4 K (26{degrees}F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175{degrees}F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6{degrees}F).

  16. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core; Determinacion de la fluencia neutronica en las soldaduras del 'core shroud' del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C. [ININ, 52750 Ocoyoacac, Estado de mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2006-07-01

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

  17. A new magnetotelluric monitoring network operating in Agri Valley (Southern Italy: study of stability of apparent resistivity estimates

    Directory of Open Access Journals (Sweden)

    L. Telesca

    2008-06-01

    Full Text Available Variations detected in geophysical, especially electromagnetic, parameters in seismic active areas have been sometimes attributed to modifications of the stress field. Among the different geophysical methods, magnetotellurics (MT could be one of the most effective because it allows us to explore down to seismogenic depths. Continuous MT recording could allow us to evaluate whether possible variations are significantly correlated with the seismic activity of investigated area. To assess the significance of such observations we must be able to say how well an apparent resistivity curve should be reproduced when measurements are repeated at a later time. To do this properly it is essential to know that the estimated error bars accurately represent the true uncertainties in comparing the transfer functions. In this work we will show the preliminary results obtained from the analysis of the data coming from the new MT monitoring network installed in Agri Valley. This analysis gives us the possibility: i to better study the temporal stability of the signals, ii to better discriminate the noise affecting the measures by remote reference estimation. The performed analysis disclosed a relatively low degree of noise in the investigated area, which is a promising condition for monitoring.

  18. CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy (Finland); Bjoere, S.; Olsson, Lena [ABB Atom AB, Vaesteraas (Sweden)

    1992-12-01

    Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems.

  19. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, Carl, E-mail: carl.adamsson@psi.ch [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden); Le Corre, Jean-Marie, E-mail: lecorrjm@westinghouse.com [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden)

    2011-08-15

    Highlights: > The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. > A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. > MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. > The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. > The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle

  20. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  1. Hardiness as a predictor of mental health and well-being of Australian army reservists on and after stability operations.

    Science.gov (United States)

    Orme, Geoffrey J; Kehoe, E James

    2014-04-01

    This study tested whether cognitive hardiness moderates the adverse effects of deployment-related stressors on health and well-being of soldiers on short-tour (4-7 months), peacekeeping operations. Australian Army reservists (N = 448) were surveyed at the start, end, and up to 24 months after serving as peacekeepers in Timor-Leste or the Solomon Islands. They retained sound mental health throughout (Kessler 10, Post-Traumatic Checklist-Civilian, Depression Anxiety Stress Scale 42). Ratings of either traumatic or nontraumatic stress were low. Despite range restrictions, scores on the Cognitive Hardiness Scale moderated the relationship between deployment stressors and a composite measure of psychological distress. Scatterplots revealed an asymmetric pattern for hardiness scores and measures of psychological distress. When hardiness scores were low, psychological distress scores were widely dispersed. However, when hardiness scores were higher, psychological distress scores became concentrated at a uniformly low level.

  2. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  3. Nuclear-coupled thermal-hydraulic stability analysis of boiling water reactors

    Science.gov (United States)

    Karve, Atul A.

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model we developed from: the space-time modal neutron kinetics equations based on spatial omega-modes, the equations for two-phase flow in parallel boiling channels, the fuel rod heat conduction equations, and a simple model for the recirculation loop. The model is represented as a dynamical system comprised of time-dependent nonlinear ordinary differential equations, and it is studied using stability analysis, modern bifurcation theory, and numerical simulations. We first determine the stability boundary (SB) in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value and then transform the SB to the practical power-flow map. Using this SB, we show that the normal operating point at 100% power is very stable, stability of points on the 100% rod line decreases as the flow rate is reduced, and that points are least stable in the low-flow/high-power region. We also determine the SB when the modal kinetics is replaced by simple point reactor kinetics and show that the first harmonic mode has no significant effect on the SB. Later we carry out the relevant numerical simulations where we first show that the Hopf bifurcation, that occurs as a parameter is varied across the SB is subcritical, and that, in the important low-flow/high-power region, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line. Hence, a point on the 100% rod line in the low-flow/high-power region, although stable, may nevertheless be a point at which a BWR should not be operated. Numerical simulations are then done to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is determined that the NRC requirement of DR loop model that we develop is studied by carrying

  4. Development and validation of advanced CFD models for detailed predictions of void distribution in a BWR bundle

    Science.gov (United States)

    Neykov, Boyan

    In recent years, a commonly adopted approach is to use Computational Fluid Dynamics (CFD) codes as computational tools for simulation of different aspects of the nuclear reactor thermal-hydraulic performance where high-resolution and high-fidelity modeling is needed. Within the framework of this PhD work, the CFD code STAR-CD [1] is used for investigations of two phase flow in air-water systems as well as boiling phenomena in simple pipe geometry and in a Boiling Water Reactor (BWR) fuel assembly. Based on the two-fluid Eulerian solver, improvements of the STAR-CD code in the treatment of the drag, lift and wall lubrication forces in a dispersed two phase flow at high vapor (gas) phase fractions are investigated and introduced. These improvements constitute a new two phase modeling framework for STAR-CD, which has been shown to be superior as compared to the default models in STAR-CD. The conservation equations are discretized using the finite-volume method and solved using a solution procedure is based on Pressure Implicit with Splitting of Operators (PISO) algorithm, adapted to the solution of the two-fluid model. The improvements in the drag force modeling include investigation and integration of models with dependence on both void fraction and bubble diameter. The set of the models incorporated into STAR-CD is selected based on an extensive literature review focused on two phase systems with high vapor fractions. The research related to the modeling of wall lubrication force is focused on the validation of the already existing model in STAR-CD. The major contribution of this research is the development and implementation of an improved correlation for the lift coefficient used in the lift force formula. While a variety of correlations for the lift coefficient can be found in the open literature, most of those were derived from experiments conducted at low vapor (gas) phase fractions and are not applicable to the flow conditions existing in the BWRs. Therefore

  5. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  6. Influence of hydrogen temperature on the stability of a rocket engine combustor operated with hydrogen and oxygen

    Science.gov (United States)

    Gröning, Stefan; Hardi, Justin; Suslov, Dmitry; Oschwald, Michael

    2017-03-01

    Since the late 1960s, low hydrogen injection temperature is known to have a destabilising effect on rocket engines with the propellant combination hydrogen/oxygen. Self-excited combustion instabilities of the first tangential mode have been found recently in a research rocket combustor operated with the propellant combination hydrogen/oxygen with a hydrogen temperature of 95 K. A hydrogen temperature ramping experiment has been performed with this research combustor to analyse the impact of hydrogen temperature on the self-excited combustion instabilities. The temperature was varied between 40 and 135 K. Contrary to past results found in literature, the combustor was found to be stable at low hydrogen temperatures while increased oscillation amplitudes of the first tangential mode were found at higher temperatures of around 100 K and above, which is consistent with previous observations of instabilities in this combustor. Further analysis shows that hydrogen temperature has a strong impact on the combustion chamber resonance frequencies. By varying the hydrogen injection temperature, the frequency of the first tangential mode is shifted to coincide with the second longitudinal resonance frequency of the liquid oxygen injector. Excitation of combustion chamber pressure oscillations was observed during such events.

  7. Antioxidant (A-tocopherol acetate) effect on oxidation stability and NOx emission reduction in methyl ester of Annona oil operated diesel engine

    Science.gov (United States)

    Senthil, R.; Silambarasan, R.; Pranesh, G.

    2016-11-01

    There is a major drawback while using biodiesel as a alternate fuel for compression ignition diesel engine due to lower heating value, higher viscosity, higher density and higher oxides of nitrogen emission. To minimize these drawbacks, fuel additives can contribute towards engine performance and exhaust emission reduction either directly or indirectly. In this current work, the test was conducted to investigate the effect of antioxidant additive (A-tocopherol acetate) on oxidation stability and NOx emission in a of Annona methyl ester oil (MEAO) fueled diesel engine. The A-tocopherol acetate is mixed in different concentrations such as 0.01, 0.02, 0.03 and 0.04% with 100% by vol MEAO. It is concluded that the antioxidant additive very effective in increasing the oxidation stability and in controlling the NOx emission. Further, the addition of antioxidant additive is slight increase the HC, CO and smoke emissions. Hence, A-tocopherol acetate is very effective in controlling the NOx emission with MEAO operated diesel engine without any major modification.

  8. Remarks on boiling water reactor stability analysis. Pt. 1. Theory and application of bifurcation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lange, Carsten; Hurtado, Antonio [Technische Univ. Dresden (Germany). Chair of Hydrogen and Nuclear Energy; Schuster, Roland [Kernkraftwerk Brunsbuettel GmbH und Co. oHG, Brunsbuettel (Germany); Lukas, Bernard [EnBW Kernkraft GmbH, Philippsburg (Germany). Kernkraftwerk Philippsburg; Aguirre, Carlos [Kernkraftwerk Leibstadt AG, Aargau (Switzerland); Hennig, Dieter

    2012-11-15

    Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants. (orig.)

  9. Maintain Stability Operations Capability During Military Operations

    Science.gov (United States)

    2013-03-01

    hegemony. Rising states, primarily identified as the BRIC countries, (Brazil, Russia, India and China), but potentially including South Africa, Korea...occurred in cooperation and integration between departments and agency due to Secretary-level cooperation. Particularly innovative was then-Secretary...Without innovative solutions, current programming and training strategies will be hard-pressed to eliminate “preparation-execution” training gaps or

  10. Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

  11. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Wagner, K.C. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  12. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  13. Increment of the operational stability of the NGL I plant located in the Tablazo Petrochemical Complex, working with a low feed gas flow rate

    Energy Technology Data Exchange (ETDEWEB)

    Mendez, Alejandro

    2004-07-01

    The function of the plant NGL I is to extract the ethane and the liquids of the current of natural gas coming from Maracaibo Lake, emphatically in the ethane and propane for being a raw material for the olefins synthesis. The needed temperature for the condensation takes place cooling the dehydrated feed gas, by means of the heat exchange with two currents, cooling propane and the other cold methane gas. In 1998 some improvements to the plant were made, which are known as (MERE) (Ethane Recovery Improvements Project), these modifications were made in order to enhance the ethane recovery during the process, however, when it started to work under this modifications high instability in the performance of the compressor located in the refrigeration system was observed, which produced the shutdown of the ethane recovery process when working with feed gas stream flow rates (under 130 MMSCFD). Moreover, during the last years a decrease in the feed gas flow has been observed, in a mean of 120 MMSCFD, which has had a negative effect on the plant, because it has worked in a non-continuous state. Thanks to several improvements implemented in the actuality, we have achieved to diminish the minimal flow of operation of the plant to 95 MMPCSD, allowing the ethane production at this flow of feeding and increased the operational stability of the plant of LNG I. (author)

  14. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  15. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  16. Evaluation of the reduction of boron-10 in the control rods in the BWR of the Laguna Verde Central, through steady state calculations; Evaluacion de la reduccion del Boro-10 en las barras de control en los BWR de la CLV, mediante calculos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J.L.; Perusquia, R.; Hernandez, J.L.; Ramirez S, J.R. [Departamento de Sistemas Nucleares, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    One of the more important aspects related with the safety and economy in the operation of a nuclear power reactor, it is without a doubt the control of the reactivity. During the normal operation of a reactor of boiling water (BWR-Boiling Water Reactor), the control of the reactivity in the nucleus it is strongly determined by the efficiency of the control rods. In the case of the Laguna Verde Nuclear power station (CNLV) the nucleus of the reactors has 109 control rods grouped in 4 sets. The CNLV at the moment uses the CCC method (Control Cell Core) in the design of the cycle. With this method only the A2 group is used for the control of the reactivity at full power. With the purpose of quantifying the effect of the decrease of the burnable poison (B{sub 4}C) of the control rods and in particular to the effect due to the postulated lost of 10% of Boron 10, it was carried out a series of calculations of the nucleus in stationary state by means of the system of HELIOS/CM-PRESTO codes. In this work the main derived results of these 3D simulations(three dimensions) of the reactors of the CNLV are presented. It was analyzed the one behavior of the infinite neutron multiplication factor (K{sub infinite}), at fuel assemble cell level used in an equilibrium cycle for the CNLV. It was also analyzed the effect in the shutdown margin (ShutDown Margin- SDM) in cold condition CZP (Cold Zero Power). Its are also included those results of the ARI cases (All Rods In) and SRO (Strong Rod Out). From the cases in condition HFP (Hot Full Power) the behavior of the effective multiplication factor (K{sub eff}) is presented. (Author)

  17. Numerical Simulations and Design Optimization of the PHT Loop of Natural Circulation BWR

    Directory of Open Access Journals (Sweden)

    G. V. Durga Prasad

    2008-01-01

    Full Text Available Mathematical modeling and numerical simulation of natural circulation boiling water reactor (NCBWR are very important in order to study its performance for different designs and various off-design conditions and for design optimization. In the present work, parametric studies of the primary heat transport loop of NCBWR have been performed using lumped parameter models and RELAP5/MOD3.4 code. The lumped parameter models are based on the drift flux model and homogeneous equilibrium mixture (HEM model of two-phase flow. Numerical simulations are performed with both models. Compared to the results obtained from the HEM model, those obtained from the drift flux model are closer to RELAP5. The variations of critical heat flux with various geometric parameters and operating conditions are thoroughly investigated. The material required to construct the primary heat transport (PHT loop of NCBWR has been minimized using sequential quadratic programming. The stability of NCBWR has also been verified at the optimum point.

  18. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  19. Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

  20. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms; Una metodologia para obtener los patrones de barras de control en un BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Montes T, J.L. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Requena R, I. [Universidad de Granada, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

    2003-07-01

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  1. Assessing Locally Focused Stability Operations

    Science.gov (United States)

    2014-01-01

    objectives to survive: sustainability, public legitimacy, and the ability to create chaos and instability. Based on historical and fictional scenarios...environment does not mean blindly adopting the locals’ goals, but having such a dialogue can inject greater realism into the design of the

  2. Force Sizing for Stability Operations

    Science.gov (United States)

    2010-03-01

    Angola ‘88‐’99 Argentina ‘69‐’83 Bolivia ‘67 Borneo ( Malaysia ) ‘63‐’66 Cuba ‘56-59 Cyprus ‘55-’59 Dhofar (Oman) ‘65-’75 El Salvador ‘79-’92 Greece ‘46...outcome of this campaign as a failure. Sri Lanka (1983-2009) – Success. The Sri Lankan government decisively defeated the Tamil insurgency in 2009

  3. Enhancement of Power Efficiency and Stability of P3HT-Based Organic Solar Cells under Elevated Operating-Temperatures by Using a Nanocomposite Photoactive Layer

    Directory of Open Access Journals (Sweden)

    Tran Thi Thao

    2015-01-01

    Full Text Available With the aim to find out an enhanced operating-temperature range for photovoltaic device parameters, two types of the photoactive layer were prepared: poly(3-hexylthiophene (P3HT and P3HT+nc-TiO2 (PTC thin films. The enhancement obtained for the photoelectrical conversion efficiency of the composite based OSCs is attributed to the presence of nanoheterojunctions of TiO2/P3HT. For the temperature range of 30–70°C, the decrease of the open-circuit potential was compensated by an increase of the fill factor; and the increase in the short-circuit current resulted in an overall increase of the energy conversion efficiency. At elevated temperatures of 60–80°C the efficiency of the P3HT- and PTC-based cells reached a maximum value of 1.6% and 2.1%, respectively. Over this temperature range the efficiency of P3HT-based OSC decreased strongly to zero, whereas for the PTC cells it maintained a value as large as 1.2% at the temperature range of 110–140°C. The improved thermal stability of the composite-based device was attributed to the lowered thermal expansion coefficient of the nanocomposite photoactive layer.

  4. Improving Performance and Operational Stability of Porcine Interferon-α Production by Pichia pastoris with Combinational Induction Strategy of Low Temperature and Methanol/Sorbitol Co-feeding.

    Science.gov (United States)

    Gao, Min-Jie; Zhan, Xiao-Bei; Gao, Peng; Zhang, Xu; Dong, Shi-Juan; Li, Zhen; Shi, Zhong-Ping; Lin, Chi-Chung

    2015-05-01

    Various induction strategies were investigated for effective porcine interferon-α (pIFN-α) production by Pichia pastoris in a 10 L fermenter. We found that pIFN-α concentration could be significantly improved with the strategies of low-temperature induction or methanol/sorbitol co-feeding. On this basis, a combinational strategy of induction at lower temperature (20 °C) with methanol/sorbitol co-feeding has been proposed for improvement of pIFN-α production. The results reveal that maximal pIFN-α concentration and antiviral activity reach the highest level of 2.7 g/L and 1.8 × 10(7) IU/mg with the proposed induction strategy, about 1.3-2.1 folds higher than those obtained with other sub-optimal induction strategies. Metabolic analysis and online multi-variable measurement results indicate that energy metabolic enrichment is responsible for the performance enhancement of pIFN-α production, as a large amount of ATP could be simultaneously produced from both formaldehyde oxidation pathway in methanol metabolism and tricarboxylic acid (TCA) cycle in sorbitol metabolism. In addition, the proposed combinational induction strategy enables P. pastoris to be resistant to high methanol concentration (42 g/L), which conceivably occur associating with the error-prone methanol over-feeding. As a result, the proposed combinational induction strategy simultaneously increased the targeted protein concentration and operational stability leading to significant improvement of pIFN-α production.

  5. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor; Post-procesador para simulaciones del programa ORIGEN y calculo de la composicion de la actividad de un nucleo de combustible quemado por un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval V, S. [IIE, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sandoval@iie.org.mx

    2006-07-01

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

  6. Description of a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels under BWR-conditions; Beschreibung einer einhuellenden Risswachstumskurve zum Spannungsrisskorrosionsverhalten von ferritischen Reaktordruckbehaelter (RDB)-Staehlen unter Siedewasserreaktor (SWR)-Bedingungen

    Energy Technology Data Exchange (ETDEWEB)

    Bruemmer, G. [HEW, Hamburg (Germany); Hoffmann, H. [VGB-GS, Essen (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON-Kernkraft, Hannover (Germany); Widera, M. [RWE Power, Essen (Germany); Roth, A. [Framatome ANP GmbH, Erlangen (Germany)

    2002-07-01

    The inner surface of the reactor pressure vessel of BWR reactors is lined with a welded, corrosion-resistant steel liner. In an assumed case of liner rupture down to the low-alloy ferritic base material, an integrity assessment of the pressure vesssel in consideration of the effects of reactor coolant is of utmost importance, and research in this field has been going on for more than ten years now. An analysis of the available data shows that it is now possible to describe a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels in BWR conditions. Crack growth rates of a stress intensity factor corresponding to a T/4 wall defect (i.e. 25 percent of the wall thickness) are technically not relevant. This scientific finding is supported by measurements of about 450 reactor operation years of all German LWR reactor plants, none of which showed crack initiation in the reactor pressure vessel. [German] Die mediumberuehrte Innenoberflaeche des Reaktordruckbehaelters (RDB) von Siedewasserreaktoren (SWR) ist mit einer korrosionsbestaendigen austenitischen Schweissplattierung versehen. Fuer den unterstellten Fall einer bis auf den niedriglegierten, ferritischen Grundwerkstoff durchgerissenen Pattierung ist fuer die Beurteilung der Integritaet des RDB unter Beruecksichtigung der Einwirkung des Reaktorkuehlmittels die Klaerung der Frage eines korrosionsgestuetzten Risswachstums von grosser Bedeutung. Dieses Thema ist daher bereits seit mehr als 10 Jahren Gegenstand umfangreicher Forschungsaktivitaeten. Ende der 80er- und Anfang der 90er-Jahre wurden fuer ferritische RDB-Staehle von SWR-Anlagen Risswachstumsgeschwindigkeiten veroeffentlicht, die binnen weniger als einem Jahr zum Durchriss der drucktragenden Wand eines RDB gefuehrt haetten. Daraufhin wurden internationale Forschungsaktivitaeten zur Ermittlung zuverlaessiger und reproduzierbarer Risswachstumsdaten initiiert, deren Ergebnisse zusammenfassend dargestellt werden. Die

  7. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    Science.gov (United States)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  8. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  9. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

  10. Design and optimization of a fuel reload of BWR with plutonium and minor actinides; Diseno y optimizacion de una recarga de combustible de BWR con plutonio y actinidos menores

    Energy Technology Data Exchange (ETDEWEB)

    Guzman A, J. R.; Francois L, J. L.; Martin del Campo M, C.; Palomera P, M. A. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: maestro_juan_rafael@hotmail.com

    2008-07-01

    In this work is designed and optimized a pattern of fuel reload of a boiling water reactor (BWR), whose fuel is compound of uranium coming from the enrichment lines, plutonium and minor actinides (neptunium, americium, curium); obtained of the spent fuel recycling of reactors type BWR. This work is divided in two stages: in the first stage a reload pattern designs with and equilibrium cycle is reached, where the reload lot is invariant cycle to cycle. This reload pattern is gotten adjusting the plutonium content of the assembly for to reach the length of the wished cycle. Furthermore, it is necessary to increase the concentration of boron-10 in the control rods and to introduce gadolinium in some fuel rods of the assembly, in order to satisfy the margin approach of out. Some reactor parameters are presented: the axial profile of power average of the reactor core, and the axial and radial distribution of the fraction of holes, for the one reload pattern in balance. For the design of reload pattern codes HELIOS and CM-PRESTO are used. In the second stage an optimization technique based on genetic algorithms is used, along with certain obtained heuristic rules of the engineer experience, with the intention of optimizing the reload pattern obtained in the first stage. The objective function looks for to maximize the length of the reactor cycle, at the same time as that they are satisfied their limits related to the power and the reactor reactivity. Certain heuristic rules are applied in order to satisfy the recommendations of the fuel management: the strategy of the control cells core, the strategy of reload pattern of low leakage, and the symmetry of a quarter of nucleus. For the evaluation of the parameters that take part in the objective function it simulates the reactor using code CM-PRESTO. Using the technique of optimization of the genetic algorithms an energy of the cycle of 10834.5 MW d/tHM is obtained, which represents 5.5% of extra energy with respect to the

  11. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  12. Determination of the local flow in a BWR by means of neutron noise analysis; Determinacion del caudal local en un BWR por medio del analisis de ruido neutronico

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R.; Rojas S, A. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Blazquez M, J. [CIEMAT, Madrid (Spain); Calleros M, G. [CFE, Alto Lucero, Veracruz (Mexico)]. e-mail: jare@nuclear.inin.mx

    2007-07-01

    The signal to noise of the local power of range monitors (LPRM for their initials in English) its are used to estimate the reactor power, but this signal includes information of the dynamic behavior of a boiling water reactor. An interesting parameter that one can obtain by means of the data of the local power monitors is the resistance time of the holes. That to make with the measures of these times that are inversely proportional to the local flow? It can be carried out a 'local surveillance' of the flow. In this work a relationship that allows to obtain the permanency time of the holes in the refrigeration channel in the reactor core with base in the mensurations of the index of the decay ratio and of the characteristic frequency value of the reactor of the neutronic signal in stationary state. The method of the first minimum of the transfer function between two detectors of neutron flow to different height to obtain the local residence time of the holes is also presented. For the determination of the permanency time of the holes in the refrigeration channel obtained signals of the backs of the SIIP of special sampling of the Unit 2 of Laguna Verde Central during 2006 were used. Firstly its were preconditioning the power signals and the signals of the neutronic noise to which are applied the first minimum method in the transfer function were obtained. Later on the STABIL code was used which contains an autoregressive model to obtain the DR and the frequency, with those that the holes residence times were determined. (Author)

  13. One-Year Post-Operative Stability of Lefort I Osteotomies Using Resorbable Fixation: A Retrospective Analysis of Diverse Facial Patterns on Skeletal Relapse

    Science.gov (United States)

    2003-06-01

    osteosynthesis material for stabilization of midface fractures: experimental investigation in sheep . J Cranio-Maxillofac Surg 27 (1):51-57. Bailey L...1952). Treatment of developmental malformations of the jaws. Plast Reconstr Surg. 10: 473. Costa F, Robiony M, Politi M. (1999). Stability of Le Fort...fixation for congenital pediatric craniofacial surgery. J Craniofac Surg. 8 (2): 110-115. Habal M. (1997). Absorbable, invisible, and flexible

  14. Thermomechanical evaluation of BWR fuel elements for procedures of preconditioned with FEMAXI-V; Evaluacion termomecanica de elementos combustible BWR para procedimientos de preacondicionado con FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Lucatero, M.A.; Ortiz V, J. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2006-07-01

    The limitations in the burnt of the nuclear fuel usually are fixed by the one limit in the efforts to that undergo them the components of a nuclear fuel assembly. The limits defined its provide the direction to the fuel designer to reduce to the minimum the fuel failure during the operation, and they also prevent against some thermomechanical phenomena that could happen during the evolution of transitory events. Particularly, a limit value of LHGR is fixed to consider those physical phenomena that could lead to the interaction of the pellet-shirt (Pellet Cladding Interaction, PCI). This limit value it is related directly with an PCI limit that can be fixed based on experimental tests of power ramps. This way, to avoid to violate the PCI limit, the conditioning procedures of the fuel are still required for fuel elements with and without barrier. Those simulation procedures of the power ramp are carried out for the reactor operator during the starting maneuvers or of power increase like preventive measure of possible consequences in the thermomechanical behavior of the fuel. In this work, the thermomechanical behavior of two different types of fuel rods of the boiling water reactor is analyzed during the pursuit of the procedures of fuel preconditioning. Five diverse preconditioning calculations were carried out, each one with three diverse linear ramps of power increments. The starting point of the ramps was taken of the data of the cycle 8 of the unit 1 of the Laguna Verde Nucleo electric Central. The superior limit superior of the ramps it was the threshold of the lineal power in which a fuel failure could be presented by PCI, in function of the fuel burnt. The analysis was carried out with the FEMAXI-V code. (Author)

  15. Characterization of welding of AISI 304l stainless steel similar to the core encircling of a BWR reactor; Caracterizacion de soldaduras de acero inoxidable AISI 304L similares a las de la envolvente del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gachuz M, M.E.; Palacios P, F.; Robles P, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    Plates of austenitic stainless steel AISI 304l of 0.0381 m thickness were welded by means of the SMAW process according to that recommended in the Section 9 of the ASME Code, so that it was reproduced the welding process used to assemble the encircling of the core of a BWR/5 reactor similar to that of the Laguna Verde Nucleo electric plant, there being generated the necessary documentation for the qualification of the one welding procedure and of the welder. They were characterized so much the one base metal, as the welding cord by means of metallographic techniques, scanning electron microscopy, X-ray diffraction, mechanical essays and fracture mechanics. From the obtained results it highlights the presence of an area affected by the heat of up to 1.5 mm of wide and a value of fracture tenacity (J{sub IC}) to ambient temperature for the base metal of 528 KJ/m{sup 2}, which is diminished by the presence of the welding and by the increment in the temperature of the one essay. Also it was carried out an fractographic analysis of the fracture zone generated by the tenacity essays, what evidence a ductile fracture. The experimental values of resistance and tenacity are important for the study of the structural integrity of the encircling one of the core. (Author)

  16. Comparison of results for burning with BWR reactors CASMO and SCALE 6.2 (TRITON / NEWT); Comparacion de los resultados de quemado para reactores BWR con CASMO y SCALE 6.2 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

    2014-07-01

    In this paper we compare the results from two codes burned, CASMO and SCALE 6.2 (TRITON). To do this, is simulated all segments corresponding to a boiling water reactor (BWR) using both codes. In addition, to account for different working points, simulations changing the instantaneous variables, these are repeated: void fractions (6 points), fuel temperature (6 points) and control rods (two points), with a total of 72 possible combinations of different instantaneous variables for each segment. After all simulations are completed for each segment, we can reorder the obtained cross sections, as SCALE CASMO both, to create a library of compositions nemtab format. This format is accepted by the neutronic code of nodal diffusion, PARCS v2.7. Finally compares the results obtained with PARCS and with the SIMULATE3 -SIMTAB methodology to level of full reactor. Also, we have made use of the KENO-VI and MCDANCOFF modules belonging to SCALE. The first is a Monte Carlo transport code with which you can validate the value of the multiplier, the second has been used to obtain values of Dancoff factor and increase the accuracy of model SCALE. (Author)

  17. Optimization of fuel cells for BWR using Path Re linking and flexible strategies of solution;Optimizacion de celdas de combustible para BWR empleando Path Relinking y estrategias flexibles de solucion

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J. A.; Ortiz S, J. J.; Torres V, M.; Perusquia del Cueto, R., E-mail: alejandro.castillo@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2009-10-15

    In this work are presented the obtained preliminary results to design nuclear fuel cells for boiling water reactors (BWR) using new strategies. To carry out the cells design some of the used rules in the fuel administration were discarded and other were implemented. The above-mentioned with the idea of making a comparative analysis between the used rules and those implemented here, under the hypothesis that it can be possible to design nuclear fuel cells without using all the used rules and executing the security restrictions that are imposed in these cases. To evaluate the quality of the obtained cells it was taken into account the power pick factor and the infinite multiplication factor, in the same sense, to evaluate the proposed configurations and to obtain the mentioned parameters was used the CASMO-4 code. To optimize the design it is uses the combinatorial optimization technique named Path Re linking and the Dispersed Search as local search method. The preliminary results show that it is possible to implement new strategies for the cells design of nuclear fuel following new rules. (Author)

  18. 西藏重大项目社会稳定风险评估运行框架研究%Research on the Operational Framework of Social Stability Risk Assessment of Major Projects in Tibet

    Institute of Scientific and Technical Information of China (English)

    王娟丽

    2016-01-01

    社会稳定风险评估机制将社会治理的"关口前移",预防社会稳定风险演变为危机事件,这是维稳工作的重大创新.当前,西藏初步形成了由社会稳定风险评估主体、评估范围与内容、评估程序等内容构成的社会稳定风险评估运行框架,这对于西藏稳评实践具有重要的指导意义.论文运用实例证明了此运行框架的可行性,但是由于稳评这项工作还处于探索阶段,因而此运行框架还存在不少问题,需要在实践过程中不断加强完善.%The social stability risk assessment system takes the "stage-gate antedisplacement" of social governance to prevent social stability risk evolute to the crisis event. It is a major innovation for the stability maintenance work. At present, Tibet initially formes the operational evaluation framework of social stability risk formed by the main risk assessment body, assessment scope and content, assessment process. It has an important guiding significance to the stability evaluation practice in Tibet. This paper uses instance to prove the feasibility of this framework. But the stability evaluation is still in exploratory stage, so this operational framework hs many problems, and it needs to be strengthened and perfected in the practice process.

  19. Radial distribution of UO{sub 2} and Gd{sub 2}O{sub 3} in fuel cells of a BWR Reactor; Distribucion radial de UO{sub 2} y Gd{sub 2}O{sub 3} en celdas de combustible de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia del C, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62500 (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2008-07-01

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO{sub 2} enriched in U{sup 2}35, and of the Gadolinium (Gd{sub 2}O{sub 3}). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U{sup 2}35, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U{sup 2}35, some of these rods contain a concentration of Gd{sub 2}O{sub 3} and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

  20. Controlling the feedwater flow in a BWR. Examples from Forsmark 2; Regleringen av matarvattenfloedet i en BWR. Med exempel fraan Forsmark 2

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, Bengt-Goeran; Oguma, Ritsuo (GSE Power Systems AB, Nykoeping (Sweden))

    2009-03-15

    An investigation of the feedwater controller at Forsmark 2 has been performed. The investigation is based on signal analysis of measurement signals recorded during operation of the plant during different tests. The feedwater controller consists of the water level controller, the flow controller and the condenser balance controller. The overall goal of the feedwater control is to maintain constant water level (level controller) in the reactor and at the same time balance the water levels in the two condensers (condenser balance controller) to avoid that one condenser is full of water while the other one is operated with too low level. There is also a feed forward of the difference between steam flow and feedwater flow (flow controller) for each turbine system with the aim to reduce the fluctuation in reactor water level. The relation in strength between the three controllers is such that the level controller is the strongest followed by the condenser balance controller and finally the flow controller. Tests with trip of the feedwater pump and automatic start of the spare pump in each turbine system indicates a fast reduction in reactor water level that is restored after the transient in the control system. The transient in water level is stable without oscillations. However, it takes about 100 s before the reactor water level is restored. The function of the flow controller has been questioned by the authors. It does not take the action that is expected when a disturbance takes place in the difference between steam and feedwater flow. In addition to this principal weakness there is an offset in the feedwater controller output for feedwater flow 22 that reduces the contribution in flow control that is expected during the introduction of a disturbance. This offset should be adjusted during instrument maintenance of the feedwater controller. The PIP parameters for the level controller are gain factors and time constants. These have been evaluated with the aid of

  1. Radiological consequence assessments of degraded core accident scenarios derived from a generic Level 2 PSA of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu; Ishikawa, Jun; Tomita, Kenichi; Muramatsu, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-12-01

    The radiological consequence assessments have been made of postulated core damage accidents with source terms derived from a generic Level 2 PSA of a BWR carried out by the Japan Atomic Energy Research Institute (JAERI). The source terms used were for the five core damage accident sequences with the drywell and wetwell failure cases, the release control case by venting of the containment and the accident termination case by the containment spray. The radiological consequences have been assessed for individual dose, collective dose, individual risk of early health effects and individual risk of late health effects by a probabilistic accident consequence assessment code, OSCAAR developed in JAERI. Following conclusions were obtained for the assumed source terms. In case of the over pressure failures of the primary containment vessel, the early fatalities can be mitigated through the implementation of early countermeasures, and the late cancer fatalities remains small. For the release control and accident termination cases, the individual and collective doses to the public can be reduced without any countermeasures due to the release reduction of the volatile radionuclides such as iodine and cesium. (author)

  2. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  3. Study and characterization of noble metal deposits on similar rusty surfaces to those of the reactor U-1 type BWR of nuclear power station of Laguna Verde; Estudio y caracterizacion de depositos de metales nobles sobre superficies oxidadas similares a las del reactor de la Central de Laguna Verde (CNLV) U1 del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Flores S, V. H.

    2011-07-01

    In the present investigation work, were determined the parameters to simulate the conditions of internal oxidation reactor circulation pipes of the nuclear power plant of Laguna Verde in Veracruz. We used 304l stainless steel cylinders with two faces prepared with abrasive paper of No. 600, with the finality to obtain similar surface to the internal circulation piping nuclear reactor. Oxides was formed within an autoclave (Autoclave MEX-02 unit B), which is a device that simulates the working conditions of the nuclear reactor, but without radiation generated by the fission reaction within the reactor. The oxidation conditions were a temperature of 280 C and pressure of 8 MPa, similar conditions to the reactor operating in nuclear power plant of Laguna Verde in Veracruz, Mexico (BWR conditions), with an average conductivity of 4.58 ms / cm and 2352 ppb oxygen to simulate normal water chemistry NWC. Were obtained deposits of noble metal oxides formed on 304l stainless steel samples, in a 250 ml autoclave at a temperature range of 180 to 200 C. The elements that were used to deposit platinum-rhodium (Pt-Rh) with aqueous Na{sub 2}Pt (OH){sub 6} and Na{sub 3}Rh (NO{sub 2}){sub 6}, Silver (Ag) with an aqueous solution of AgNO{sub 3}, zirconium (Zr) with aqueous Zr O (NO{sub 3}) and ZrO{sub 2}, and zinc (Zn) in aqueous solution of Zn (NO{sub 3}){sub 2} under conditions of normal water chemistry. Also there was the oxidation of 304l stainless steel specimens in normal water chemistry with a solution of Zinc (Zn) (NWC + Zn). Oxidation of the specimens in water chemistry with a solution of zinc (Zn + NWC) was prepared in two ways: within the MEX-02 autoclave unit A in a solution of zinc and a flask at constant temperature in zinc solution. The oxides formed and deposits were characterized by scanning electron microscopy, energy dispersive X-ray analysis, elemental field analysis and X-ray diffraction. By other hand was evaluated the electrochemical behavior of the oxides

  4. Oral clonidine: a simple yet effective and safe premedicant for hemodynamic stability during laparoscopic surgery and a calm post-operative period

    Directory of Open Access Journals (Sweden)

    Ketaki Marodkar

    2016-04-01

    Conclusions: Oral clonidine in the present dose is able to maintain stable intra-operative haemodynamics and achieve a calm post-operative period during laparoscopic surgeries in ASA I/II patients. [Int J Basic Clin Pharmacol 2016; 5(2.000: 293-299

  5. Implementation of a Newton-Krylov iterative method to address strong non-linear feedback effects in FORMOSA-B BWR core simulator

    Science.gov (United States)

    Kastanya, Doddy Febrian

    A Newton-BICGSTAB solver has been developed to reduce the CPU execution time of the FORMOSA-B boiling water reactor (BWR) core simulator. The new solver treats the strong non-linearities in the problem explicitly using the Newton's method, replacing the traditionally used nested iterative approach. Taking advantage of the higher convergence rate provided by the Newton's method, assuming that a good initial estimate of the unknowns is provided, and utilizing an efficient preconditioned BICGSTAB solver, we have developed a computationally efficient Newton-BICGSTAB solver to evaluate the three-dimensional, two-group neutron diffusion equations coupled with a two-phase flow model within a BWR core simulator. The robustness of the solver has been tested against numerous BWR core configurations and consistent results have been observed each time. The best exact Newton-BICGSTAB solver performance provides an overall speedup of 2.07 to the core simulator, with reference to the traditional approach, i.e. outer (fission-source)-inner (red/black line SOR). When solving the same problem using the traditional approach but with the BICGSTAB solver as the inner iteration solver [traditional (BICGSTAB)], we observed a speedup of 1.85. This means that the Newton-BICGSTAB solver provides an additional 12% increase in the overall speedup over the traditional (BICGSTAB) solver. However, one needs to note that, on average, the exact Newton-BICGSTAB solver provides an overall speedup of around 1.70; whereas, on average, the traditional (BICGSTAB) provides an overall speedup of around 1.60. An investigation on the feasibility of implementing an inexact Newton-BICGSTAB solver indicates that further reduction in the execution time can likely be obtained through this approach. This study shows that the inexact Newton-BICGSTAB solver can provide speedups of 1.73 to 2.10 with respect to the traditional solver.

  6. Methodology for evaluation of the stability of electric systems of offshore oil platforms in interconnected operations; Metodologia para a avaliacao dos sistemas eletricos das plataformas de petroleo offshore em operacao interligada

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Dalmo Junior; Mendes, Pedro Paulo de Carvalho [Universidade Federal de Itajuba (UNIFEI), MG (Brazil)

    2004-07-01

    The electric system of the platforms usually operates in an isolated way, requesting therefore a redundancy in the generating units so that a can have a reliable and uninterrupted system of their loads. A way to improve the quality in operation in terms of safety, reliability and economy can be the connection of the platforms of petroleum that are close, since it is viable. The objective of this work is to show the methodology for the stability study of the electric system in interlinked operation in way to detail the data of the equipment that are necessary, as: One line Diagram; Transmission lines; Transformers; Power system compensator; Generating; Speed regulators; Voltage regulators; Turbines and other. Another focus of the work is to show some control models and regulation of the electric system to maintain it stable and to show models for the interconnection of two or more platforms. (author)

  7. Influence of the wet-well nodalization of a BWR3 Mark I on the containment thermal-hydraulic response during an SBO accident

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E., E-mail: luisen.herranz@ciemat.es; Fontanet, Joan; Fernández, Elena; López, Claudia

    2015-12-15

    Highlights: • Analysis of SBO sequences in BWR3 Mark I containments. • Multiple-mesh nodalization allows pool stratification set up. • Mass, momentum and energy exchanges between nodes play a key role. • Validation/verification against a scaled-down database required to credit meshing schemes. - Abstract: In the field of severe accidents simulation one of the most challenging issues is nodalization. This paper explores the effect of the wet well modeling on significant variables describing the sequence evolution. The code used for the study has been MELCOR 2.1 and the scenario chosen has been a prolonged SBO occurring in a BWR3 Mark I. The results indicate that some significant magnitudes show a moderate scatter depending on WW nodalization (i.e., core uncovery, RPV failure, hydrogen production), whereas the SP thermal state might display outstanding deviations, which sometimes affect significantly key variables like containment pressure. The difficulties and uncertainties around defining a suitable WW nodalization have been highlighted and the need to properly balance the entire plant meshing has been stressed. Even though a number of noding schemes has been explored, the results discussion underlines the importance of having a deep understanding of the potential phenomena governing the scenario and of mastering the code facilities to better model it. Some insights into WW nodalization in MELCOR 2.1 have been gained for the specific scenario (i.e., a prolonged SBO in a BWR3 Mark I) explored: a single node assumption might underestimate PCV pressurization; a loose coupling of water and gas exchanges in the WW nodalization would be preferred if the drift flux model is chosen for momentum exchange in the flow pathways between WW nodes; the potential of some axial thermal stratification in the pool should be taken into account when noding the WW; sensitivity analyses on physically supported WW nodalization schemes should be conducted and focused on key

  8. Inlet throttling effect on the boiling two-phase flow stability in a natural circulation loop with a chimney

    Science.gov (United States)

    Furuya, M.; Inada, F.; Yasuo, A.

    Experiments have been conducted to investigate an effect of inlet restriction on the thermal-hydraulic stability. A Test facility used in this study was designed and constructed to have non-dimensional values that are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation at the stability boundary was described as a function of heat flux and inlet subcooling independent of inlet restriction. In order to extend experimental database regarding thermal-hydraulic stability to different inlet restriction, numerical analysis was carried out based on the homogeneous flow model. Stability maps in reference to the core inlet subcooling and heat flux were presented for various inlet restrictions using the above-mentioned function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux.

  9. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  10. Last experiences on ID BWR shroud inspection and the new developments to examine the below core plate areas

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.; Willke, A.; Yague, L. [TECNATOM SA, Madrid (Spain)

    2001-07-01

    In recent years, the owners of BWR type nuclear power plants have had to address new inspection requirements relating to the core shroud inside the reactor vessel, the aim of which is to contain the fuel assemblies and provide support for the structures located in the upper part of the reactor. The shroud consists of a cylinder measuring some 40-50 mm in thickness, manufactured from various sections of AISI-304 stainless steel and INCONEL, joined by vertical and circumferential welds. The appearance of unstable cracks in these welds would directly affect the structural integrity of the component and the safety of the plant. As regards access to the core shroud and to the surface to be examined, two alternatives might be considered: inspection from outside the component, moving along the so-called annulus between the reactor vessel wall and the component (OD inspection), or from the interior (ID inspection). With a view to addressing this problem, Tecnatom has in recent years launched several projects, grouped under the generic name TEIDE, in order to develop scanners and NDT techniques achieving the maximum inspection coverage of this component. The decision was taken to perform ID inspections, mainly because this type of scanners were not available at that time, and which provide the 4 following advantages. 1) Maximum inspected weld length. This avoids interference with the jet pumps and the systems present in the annulus and affecting OD inspections. Besides, the repairs performed on in-service core shrouds in all cases imply the addition of new fixed elements on their outer surface, since the fuel assembly space must be left free. 2) Reduction of inspection times and of unforeseen events: maintenance of planning schedules, reduction of personnel doses, reduced critical path time. 3) High inspection accuracy and repeatability. 4) Simplification of equipment positioning work (similar to the installation of fuel assemblies). As regards inspection techniques, the

  11. Experimental data report for test TS-2; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1993-01-01

    本報告書は、1990年2月に実施した照射済BWR燃料を用いた2回目の反応度事故模擬実験であるTS-2について実験データをまとめたものである。TS-2実験に使用した試験燃料は初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3Gwd/tであった。NSRRにおける照射実験は、大気圧、室温の静止水冷却条件下で行い、発熱量は72pm5cal/g・fuel(ピークエンタルピ66pm5cal/g・fuel)を与えた。その結果燃料破損は生じなかった。実験条件、実験方法、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  12. Experimental data report for test TS-1; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1992-01-01

    本報告書は、1989年10月に実施した照射済BWR燃料を用いた最初の反応度事故模擬実験であるTS-1について、実験データをまとめたものである。TS-1実験に使用した試験燃料は、初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3GWd/tであった。NSRRにおける照射実験は、新たに開発した専用の2重カプセルを用い、大気圧・室温の静止水冷却条件下で行い、発熱量61cal/g・fuel(ピークエンタルピ55cal/g・fuel)を与えた。その結果、燃料破損は生じなかった。実験条件、実験方法、燃料燃焼度の測定結果、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  13. Validation of the CASMO-4 code against SIMS-measured spatial gadolinium distributions inside a BWR pin

    Energy Technology Data Exchange (ETDEWEB)

    Holzgrewe, F.; Gavillet, D.; Restani, R.; Zimmermann, M.A

    2000-07-01

    The purpose of the present study was to establish a database, useful for the assessment of the predictive capabilities of assembly burnup codes with respect to the depletion of the burnable absorber gadolinium (Gd). An SVEA-96 fuel assembly containing one unique Gd rod, with an initial Gd{sub 2}O{sub 3}-content of 9 wt%, was irradiated for one cycle in a Swiss Boiling Water Reactor (BWR), and then transported to the PSI hotcells for post-irradiation examination. Relative radial and azimuthal Gd distributions were obtained from Secondary Ion Mass Spectrometry (SIMS) at three axial positions. Two perpendicular line scans were performed at each position in order to capture the expected asymmetry in the Gd depletion. Since such high-spatial-resolution experimental data for individual fuel pins are quite rare, they form a valuable basis for the further validation of the calculational methods in reactor physics codes. The goal of this study was to contribute to the validation of the micro-region depletion model of CASMO-4 with respect to its standard application of generating two-group cross sections for the 3-D core simulator SIMULATE-3. The only notable difference to the standard application is a more detailed noding scheme for the Gd pin, required to obtain an improved resolution of the calculated distributions. The comparison of measurements with calculational results was found to be quite insensitive to the axial position, and the agreement was found to be very good for all isotopes investigated. The two important neutron-absorbing isotopes {sup 155} Gd and {sup 157} Gd, in particular, show excellent agreement. In conclusion, the CASMO-4 micro-region depletion model has been demonstrated to accurately predict the evolution of the radial distribution of the burnable absorber gadolinium. (authors)

  14. Sol-gel encapsulation of pullulanase in the presence of hybrid magnetic (Fe3O4-chitosan) nanoparticles improves thermal and operational stability.

    Science.gov (United States)

    Long, Jie; Li, Xingfei; Zhan, Xiaobei; Xu, Xueming; Tian, Yaoqi; Xie, Zhengjun; Jin, Zhengyu

    2017-02-27

    Pullulanase was sol-gel encapsulated in the presence of magnetic chitosan/Fe3O4 nanoparticles. The resulting immobilized pullulanase was characterized by scanning electron microscopy, vibrating sample magnetometry, Fourier transform infrared spectroscopy and thermogravimetric analysis. The results showed that the addition of pullulanase created a more regular surface on the sol-gel matrix and an enhanced magnetic response to an applied magnetic field. The maximal activity retention (83.9%) and specific activity (291.7 U/mg) of the immobilized pullulanase were observed under optimized conditions including an octyltriethoxysilane:tetraethoxysilane (OTES:TEOS) ratio of 1:2 and enzyme concentration of 0.484 mg/mL sol. The immobilized enzyme exhibited good thermal stability. When the temperature was above 60 °C, the immobilized pullulanase showed significantly higher activity than the free enzyme (p sol-gel encapsulation and co-immobilized by crosslinking-encapsulation retained 52 and 69% of their initial activity after 5 h at 62 °C, respectively, compared to 11% for the free enzyme. Moreover, the stability of the pullulanase was improved by crosslinking-encapsulation, as the enzyme retained more than 85 and 81% of its original activity after 5 and 6 consecutive reuses, respectively, compared to 80 and 72% of its original activity for simple sol-gel encapsulated enzymes. This indicated the leakage of enzyme molecules through the pores of the gel was substantially abated by cross-linking. Such immobilized pullulanase provides high stability and ease of enzyme recovery, characteristics that are advantageous for applications in the food industry that involve continuous starch processing.

  15. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

  16. Power-flow control and stability enhancement of four parallel-operated offshore wind farms using a line-commutated HVDC link

    DEFF Research Database (Denmark)

    Wang, Li; Wang, Kuo-Hua; Lee, Wei-Jen

    2010-01-01

    This paper presents an effective control scheme using a line-commutated high-voltage direct-current (HVDC) link with a designed rectifier current regulator (RCR) to simultaneously perform both power-fluctuation mitigation and damping improvement of four parallel-operated 80-MW offshore wind farms...

  17. Definition of loss-of-coolant accident radiation source: summary and conclusions. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.; Lurie, N.A.; Houston, D.H.; Naber, J.A.

    1978-05-01

    The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as time-dependent beta and gamma-ray energy release rates and spectra which are fundamental quantities that are not specific to a plant design but are generally applicable to any nuclear power station.

  18. Preliminary planning study for safety relief valve experiments in a Mark III BWR pressure suppression system

    Energy Technology Data Exchange (ETDEWEB)

    McCauley, E.W.; Holman, G.S.

    1980-04-21

    In response to a request from the Water Reactor Safety Research Division of the US NRC, a preliminary study is provided which identifies key features and consideration involved in planning a comprehensive in-plant Safety Relief Valve experimental program for a Mark III containment design. The report provides identification of program objectives, measurement system requirements, and some details quantifying expected system response. In addition, a preliminary test matrix is outlined which involves a supporting philosophy intended to enhance the usefulness of the experimental results for all members of the program team: experimentalists, analysts, and plant operator.

  19. Physical protection of nuclear facilities. Quarterly progress report, July--September 1978. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, L.D. (ed.)

    1979-01-01

    Major activities during the fourth quarter of FY78 included (1) the vital area analysis of operational reactors and characterization of the Standardized Nuclear Unit Power Plant System (SNUPPS), (2) the algorithm development of a new pathfinding computer code, (3) the completion of contractor-supported work for the component generic data base, (4) the refinement of tests related to human parameters modeling, and (5) the addition of improvements to and demonstration of the Safeguards Automated Facility Evaluation (SAFE), Safeguards Network Analysis Procedure (SNAP), and Fixed-Site Neutralization Model (FSNM) methodologies.

  20. Development of large capacity main steam safety relief valve in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hama, Teruo; Takasaki, Kiyoshi [Electric Power Development Co. Ltd., Tokyo (Japan); Takeshima, Kikuo; Hamada, Nobuyoshi

    2000-08-01

    As a change of reactor performance on design of an improved type boiling water reactor aiming at loading of uranium-plutonium mixed oxides fuel (MOX fuel) at Oma town in Aomori prefecture, negative absolute value in void coefficient has an increasing trend due to loading of the MOX fuel. As a result, pressure increase in reactor pressure vessel has increasing trend at forming of pressurized transient phenomenon such as negative loading loss, and then it was attempted to increase total flushing amount of a main stream safety relief valve (MSSRV) for its relaxation. Then, in order to control increase of setting numbers of the valve, relax limitation of arrangement of its storing vessel, and control increase of maintenance and inspection works, a large capacity MSSRV of 460 t/h with about 16% larger flushing amount than that of present MSSRV was planned to an actual reactor. As a result, it could be confirmed by its functional tests and so forth that it had nearly equal correlation of flushing pressure on nitrogen and steam to that of the present MSSRV, nearly same reliability on enough durability, good working stability, and so forth as that of the present MSSRV, and no obstacle on its application to an actual reactor. (G.K.)

  1. Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2007-07-01

    Presently work the {beta} and {gamma} radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of {beta} and {gamma} radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the {gamma} radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled

  2. IA-D3: An Assessment Process for Military Health Support in Stability Operations (MSOs), Enabling Medical Diplomacy for Countering Irregular Threats

    Science.gov (United States)

    2011-05-05

    health, pediatrics , tropical medicine, geriatrics])?. (2) Do medical persom1el have training or experience in CMO (language/cultural skills, civil...34Developing Future Command Surgeons and Staff for Joint Operations and Assignments." JCOA Journal (Summer 2010): 51. 34 APPENDIX A - HSSChecklist...health, pediatrics , adequate number of female providers, etc.)? (3) .Will other MNF (multinational force) nations conduct or support medical civil

  3. Experimental facility for containment sump reliability studies (Generic Task A-43). [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Durgin, W. W.; Padmanabhan, M.; Janik, C. R.

    1980-12-01

    On July 3, 1979, Sandia National Laboratories (Sandia) contracted the Alden Research Laboratory (ARL) to conduct tests on unresolved safety issues associated with containment sump performance during the recirculation mode (Generic Task A-43). This report describes the test facility constructed and completed under Phase I, Task III of the contract. Sump performance is determined through the observation of vortex formation in the main tank and the measurement of swirl, pressure gradient, and entrained air in the suction pipes. The use of electrically operated valves and a sophisticated data acquisition system, with computer interface, allows the test flow parameters to be set and test data to be taken (with the exception of vortex observations) from a single central office.

  4. Refined multi-level methodology in parallel computing environment for BWR RIA analysis

    Science.gov (United States)

    Solis-Rodarte, Jorge

    2000-12-01

    Best-estimate methodologies in boiling water reactor can reduce the traditional conservative thermal margins imposed on the designs and during the operation of this type of nuclear reactors. Traditional operating thermal margins are obtained based on simplified modeling techniques that are supplemented with the required dose of conservatism. For instance, much more realistic transient pin peaking distributions can be predicted by applying a dehomogenization algorithm, based on a flux reconstruction scheme which uses nodal results during both steady state and transient calculation at each time step. A subchannel analysis module for obtaining thermal margins supplements the calculation approach used. A multi-level methodology to extend the TRAC-BF1/NEM coupled code capability to obtain the transient fuel rod response has been implemented. To fulfill the development requirements some improved neutronic models were implemented into the NEM solution algorithm, namely the pin power reconstruction capability, and the simulation of a dynamic scram. The obtained results were coupled to a subchannel analysis module: COBRA-TF T-H subchannel analysis code. The objective of the pin power reconstruction capability of NEM is to locate the most limiting node (axial region of assembly/channel) within the core. The power information obtained from the NEM 3D neutronic calculation is used by the hot channel analysis module (COBRA-TF). COBRA-TF needs also the T-H conditions at the boundary nodes. This information is provided by TRACBF1 T-H system analysis code. The Subchannel analysis module uses this information to re-calculate the fluid, thermal and hydraulics conditions in the most limiting node (axial region of assembly/channel) within the core.

  5. Overweight truck shipments to nuclear waste repositories: legal, political, administrative and operational considerations

    Energy Technology Data Exchange (ETDEWEB)

    1986-03-01

    This report, prepared for the Chicago Operations Office and the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE), identifies and analyzes legal, political, administrative, and operational issues that could affect an OCRWM decision to develop an overweight truck cask fleet for the commercial nuclear waste repository program. It also provides information required by DOE on vehicle size-and-weight administration and regulation, pertinent to nuclear waste shipments. Current legal-weight truck casks have a payload of one pressurized-water reactor spent fuel element or two boiling-water reactor spent fuel elements (1 PWR/2 BWR). For the requirements of the 1960s and 1970s, casks were designed with massive shielding to accommodate 6-month-old spent fuel; the gross vehicle weight was limited to 73,280 pounds. Spent fuel to be moved in the 1990s will have aged five years or more. Gross vehicle weight limitation for the Interstate highway system has been increased to 80,000 pounds. These changes allow the design of 25-ton legal-weight truck casks with payloads of 2 PWR/5 BWR. These changes may also allow the development of a 40-ton overweight truck cask with a payload of 4 PWR/10 BWR. Such overweight casks will result in significantly fewer highway shipments compared with legal-weight casks, with potential reductions in transport-related repository risks and costs. These advantages must be weighed against a number of institutional issues surrounding such overweight shipments before a substantial commitment is made to develop an overweight truck cask fleet. This report discusses these issues in detail and provides recommended actions to DOE.

  6. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    Energy Technology Data Exchange (ETDEWEB)

    Foehl, J.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT) (Spain); Ernestova, M.; Zamboch, M. [Nuclear Research Inst. (NRI) (Czech Republic); Seifert, H.P.; Ritter, S. [Paul Scherrer Inst. (PSI) (Switzerland); Roth, A.; Devrient, B. [Framatome ANP GmbH (F ANP) (Germany); Ehrnsten, U. [Technical Research Centre of Finland (VTT) (Finland)

    2004-07-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  7. Experimental result of BWR post-CHF tests. Critical heat flux and post-CHF heat transfer coefficient. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwaki, Chikako [Toshiba Corp., Tokyo (Japan)

    2002-02-01

    Authors performed post-CHF experiments under wider pressure ranges of 2 MPa - 18 MPa, wider mass flux ranges of 33 kg/m{sup 2}s - 1651 kg/m{sup 2}s and wider superheat of heaters up to 500 K in comparison to experimental ranges at previous post-CHF experiments. Data on boiling transition, critical heat flux and post-CHF heat transfer coefficient were obtained. Used test section was 4x4-rod bundle with heaters, which diameter and length were the same as those of BWR nuclear fuels. As the result of the experiments, it was found that the boiling transition occurred just below several grid spacers, and that the fronts of the boiling transition region proceeded lower with increase of heated power. Heat transfer was due to nucleate boiling above grid spacers, while it was due to film boiling below grid spacers. Consequently, critical heat flux is affected on the distance from the grid spacers. Critical heat flux above the grid spacers was about 15% higher than that below the grid spacers, by comparing them under the same local condition. Heat transfer by steam turbulent flow was dominant to post-CHF heat transfer, when superheat of heaters was sufficiently high. Then, post-CHF heat transfer coefficient was predicted with heat transfer correlations for single-phase flow. On the other hand, when superhead of heaters was not sufficiently high, post-CHF heat transfer coefficient was higher than the prediction with heat transfer correlations for single-phase flow. Mass flux effect on post-CHF heat transfer coefficient was described by standardization of post-CHF heat transfer coefficient with the prediction for single-phase flow. However, pressure effect, superheat effect and effect of position were not described. Authors clarified that those effects could be described with functions of heater temperature and position. Post-CHF heat transfer coefficient was lowest just blow the grid spacers, and it increased with the lower positions. It increased by about 30% in one span of

  8. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J.V

    2002-07-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  9. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J.V

    2002-07-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  10. Networked Forces in Stability Operations: 101st Airborne Division, 3/2 and 1/25 Stryker Brigades in Northern Iraq

    Science.gov (United States)

    2007-01-01

    helped liberate field units from dependence on headquarters. Autonomy is further encouraged by the flat “intelligence architecture ” of the Stryker...force units that operated in northern Iraq. Rover III’s imagery is illustrated in Figure 3.4. Figure 3.5 describes the architecture used to control...on October 3–4, 1993. The objective of the opera- tion, “ Gothic Serpent,” was to capture senior figures in Mohammed Farrah Aidid’s Hadr Gidr clan

  11. A comparison of crud phases appearing on some Swedish BWR fuel rods using Laser Raman Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden)]|[Lulea Univ. of Technology (Sweden)

    2002-07-01

    Previous investigations showed that laser Raman spectroscopy (LRS) can be used as a phase specific analytical tool for radioactive fuel crud samples and also for details in the underlying layer of zirconium dioxide. It is relatively easy to record Raman spectra that discriminate between chemical phases for all crud oxides of interest. The method has therefore been recommended for crud investigations within the Swedish program. At ideal conditions the resolution is about 1 {mu}m, permitting detailed position determination of crud phases in the sample. Therefore LRS is a very good complement to X-ray diffraction (XRD). The methods for sample preparation and handling of radioactive crud samples for LRS turn out to be relatively simple. A detailed LRS study on fuel crud samples from Barsebaeck 2, Forsmark 2, Forsmark 3 and Ringhals 1 was performed in this work. All of those Swedish BWRs were operated at different conditions at the time of sampling. The chemistry regimes covered NWC, HWC and other variable conditions. Also different types of fuel, exposure times and sampling positions were selected. (authors)

  12. Advanced and flexible genetic algorithms for BWR fuel loading pattern optimization

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, Cecilia [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., 62550 (Mexico)], E-mail: cmcm@fi-b.unam.mx; Palomera-Perez, Miguel-Angel [Instituto de Investigaciones en Matematicas Aplicadas y Sistemas, Universidad Nacional Autonoma de Mexico, Circuito Escolar, Ciudad Universitaria, 04510 DF (Mexico)], E-mail: mapp@uxmcc2.iimas.unam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., 62550 (Mexico)], E-mail: jlfl@fi-b.unam.mx

    2009-10-15

    This work proposes advances in the implementation of a flexible genetic algorithm (GA) for fuel loading pattern optimization for Boiling Water Reactors (BWRs). In order to avoid specific implementations of genetic operators and to obtain a more flexible treatment, a binary representation of the solution was implemented; this representation had to take into account that a little change in the genotype must correspond to a little change in the phenotype. An identifier number is assigned to each assembly by means of a Gray Code of 7 bits and the solution (the loading pattern) is represented by a binary chain of 777 bits of length. Another important contribution is the use of a Fitness Function which includes a Heuristic Function and an Objective Function. The Heuristic Function which is defined to give flexibility on the application of a set of positioning rules based on knowledge, and the Objective Function that contains all the parameters which qualify the neutronic and thermal hydraulic performances of each loading pattern. Experimental results illustrating the effectiveness and flexibility of this optimization algorithm are presented and discussed.

  13. Low-timing-jitter, stretched-pulse passively mode-locked fiber laser with tunable repetition rate and high operation stability

    Science.gov (United States)

    Liu, Yuanshan; Zhang, Jian-Guo; Chen, Guofu; Zhao, Wei; Bai, Jing

    2010-09-01

    We design a low-timing-jitter, repetition-rate-tunable, stretched-pulse passively mode-locked fiber laser by using a nonlinear amplifying loop mirror (NALM), a semiconductor saturable absorber mirror (SESAM), and a tunable optical delay line in the laser configuration. Low-timing-jitter optical pulses are stably produced when a SESAM and a 0.16 m dispersion compensation fiber are employed in the laser cavity. By inserting a tunable optical delay line between NALM and SESAM, the variable repetition-rate operation of a self-starting, passively mode-locked fiber laser is successfully demonstrated over a range from 49.65 to 50.47 MHz. The experimental results show that the newly designed fiber laser can maintain the mode locking at the pumping power of 160 mW to stably generate periodic optical pulses with width less than 170 fs and timing jitter lower than 75 fs in the 1.55 µm wavelength region, when the fundamental repetition rate of the laser is continuously tuned between 49.65 and 50.47 MHz. Moreover, this fiber laser has a feature of turn-key operation with high repeatability of its fundamental repetition rate in practice.

  14. Importance of the operating pH in maintaining the stability of anoxic ammonium oxidation (anammox) activity in moving bed biofilm reactors.

    Science.gov (United States)

    Jaroszynski, L W; Cicek, N; Sparling, R; Oleszkiewicz, J A

    2011-07-01

    Two bench-scale parallel moving bed biofilm reactors (MBBR) were operated to assess pH-associated anammox activity changes during long term treatment of anaerobically digested sludge centrate pre-treated in a suspended growth partial nitrification reactor. The pH was maintained at 6.5 in reactor R1, while it was allowed to vary naturally between 7.5 and 8.1 in reactor R2. At high nitrogen loads reactor R2 had a 61% lower volumetric specific nitrogen removal rate than reactor R1. The low pH and the associated low free ammonia (FA) concentrations were found to be critical to stable anammox activity in the MBBR. Nitrite enhanced the nitrogen removal rate in the conditions of low pH, all the way up to the investigated level of 50mg NO(2)-N/L. At low FA levels nitrite concentrations up to 250 mg NO(2)-N/L did not cause inactivation of anammox consortia over a 2-days exposure time.

  15. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  16. Estimating boiling water reactor decommissioning costs. A user`s manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest Lab., Richland, WA (United States)

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  17. Calculations for a BWR Lattice with Adjacent Gadolinium Pins Using the Monte Carlo Cell Code Serpent v.1.1.7

    Directory of Open Access Journals (Sweden)

    Diego Ferraro

    2011-01-01

    Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.

  18. Effects of hydrogen water chemistry on corrosion fatigue behavior of cold-worked 304L stainless steel in simulated BWR coolant environments

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, M.F., E-mail: mfchiang@iner.gov.tw [Institute of Nuclear Energy Research, Division of Nuclear Fuels and Materials, Lungtan, Taoyuan 325, Taiwan (China); Young, M.C.; Huang, J.Y. [Institute of Nuclear Energy Research, Division of Nuclear Fuels and Materials, Lungtan, Taoyuan 325, Taiwan (China)

    2011-04-15

    Corrosion fatigue behavior of stainless steel 304L (SS304L) in a simulated BWR coolant with hydrogen injection was investigated. Hydrogen water chemistry slightly mitigated the corrosion fatigue degradation of the as-received SS304L specimens, but, on the contrary, it slightly increased the corrosion fatigue crack growth rates (CFCGRs) of the cold-worked specimens. All the CFCGR-tested specimens showed similar fracture features, except for the amounts of deposited corrosion debris. The results indicated that decreasing the oxygen concentration of water environment is not an effective measure to suppress the fatigue crack growth rate of cold-worked SS304L. The CFCGRs of the SS304L were determined by an interaction between corrosion, oxide-induced crack closure and cold work in corrosive environments. At a specific level of reduction, cold work could enhance the corrosion fatigue resistance of SS304 both in the air-saturated and HWC coolant environments.

  19. Effects of hydrogen water chemistry on corrosion fatigue behavior of cold-worked 304L stainless steel in simulated BWR coolant environments

    Science.gov (United States)

    Chiang, M. F.; Young, M. C.; Huang, J. Y.

    2011-04-01

    Corrosion fatigue behavior of stainless steel 304L (SS304L) in a simulated BWR coolant with hydrogen injection was investigated. Hydrogen water chemistry slightly mitigated the corrosion fatigue degradation of the as-received SS304L specimens, but, on the contrary, it slightly increased the corrosion fatigue crack growth rates (CFCGRs) of the cold-worked specimens. All the CFCGR-tested specimens showed similar fracture features, except for the amounts of deposited corrosion debris. The results indicated that decreasing the oxygen concentration of water environment is not an effective measure to suppress the fatigue crack growth rate of cold-worked SS304L. The CFCGRs of the SS304L were determined by an interaction between corrosion, oxide-induced crack closure and cold work in corrosive environments. At a specific level of reduction, cold work could enhance the corrosion fatigue resistance of SS304 both in the air-saturated and HWC coolant environments.

  20. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  1. Influence of design parameter on the operation stability of direct-drive permanent magnet synchronous motor%设计参数对直驱永磁同步电机运行平稳性的影响

    Institute of Scientific and Technical Information of China (English)

    雷静希; 肖文生

    2013-01-01

    To improve low speed operation stability of direct-drive top-drive permanent magnet synchronous motor(PMSM), optimum design was performed on the magnet pole. Two schemes of different pole-arc coefficient and eccentricity combinations were comparatively analyzed, giving the relation curves among the pole-arc coefficient, eccentricity, cogging torque and air gap flux density. The cogging torque, air gap flux density and the total harmonic distortion were analyzed numerically, and finally the optimal solution was obtained. From the above analysis results, the optimal solution is selected which can improve operation stability of the PMSM effectively.%  为改善直驱顶驱永磁同步电机低速运行时的平稳性,对永磁电机磁极进行了优化设计。对比分析了两种极弧系数与偏心距的组合方案,绘制出极弧系数、偏心距与齿槽转矩、气隙磁密之间的关系曲线;定量分析了齿槽转矩、气隙磁密和谐波畸变率,并得到最优解。经验证,采用合理的磁极极弧系数以及偏心距是改善直驱永磁同步电机空载运行平稳性的有效方法。

  2. Comparing the performance and operation stability of an SBR and MBBR for single-stage nitritation-anammox treating wastewater with high organic load.

    Science.gov (United States)

    Lackner, Susanne; Horn, Harald

    2013-01-01

    Single stage nitritation-anammox reactors have gained increasing attention for their application in municipal and industrial wastewater treatment. The most commonly used system in municipal reject water treatment is at present the sequencing batch reactor (SBR), the moving-bed biofilm reactor (MBBR) is the second most common. However, little is known about their applicability to industrial wastewaters with high C/N ratios. This study presents a comparative approach to evaluate the performance of these two systems by changing the influent from reject water (C:N ratio 1:1) stepwise to an industrial wastewater (C:N ratio 3:1). An intentionally induced temperature drop that led to nitrite accumulation was also tested. The results showed that the MBBR (1.9 kg-N m(-3) d(-1)) was superior to the SBR (0.5 kg-N m(-3) d(-1)) with at maximum up to four times higher volumetric nitrogen removal rates. Both systems accumulated nitrite (> 100 mg-N L(-1)) during the temperature drop from 30 degrees C to as low as 18 degrees C (MBBR) and 20 degrees C (SBR), which subsequently resulted in almost complete loss in the removal capacities. However, the previous removal rates could be re-established in both systems within approximately 40 days. In comparison, the MBBR showed the more stable and higher performance even though higher nitrite concentrations (up to 500 mg-N L(-1)) were encountered. Overall, MBBR operation and handling was also easier and the system was more robust to disturbances compared to the SBR.

  3. Research on control measures for the stability of micro-environmental indicators in clean operating department%洁净手术部微环境指标的稳定性控制措施

    Institute of Scientific and Technical Information of China (English)

    高兴莲; 沈剑辉; 赵晶; 吴荷玉; 杨英; 王曾妍; 胡娟娟

    2014-01-01

    目的:探讨洁净手术部微环境指标稳定性控制的有效措施。方法通过聘请洁净技术专业工程人员参与管理层流机组运行与维护,手术室管理者定期抽样百级、千级、万级、十万级手术间微环境中温度、湿度、压差、照明与声贝5项指标,与国家标准中手术部微环境指标参数进行对照;对洁净手术部全年沉降菌监测结果进行比较。结果洁净手术部微环境各项指标在专业技术人员管理下,百级、千级、万级、十万级手术间微环境中温度、湿度、压差、照明与声贝均在行业规定值范围内;2013年百级、千级、万级、十万级手术间沉降菌监测结果,在GB50333国标值范围内,且低于国标规定值。结论专业工程技术人员参与洁净手术部微环境管理维护,手术室管理者进行不定期抽样监督,手术部微环境各项指标可得到很好地控制。%Objective To explore the effective measures for the stability of micro-environmental indicators in clean operating department.Methods Professionals of clean technology were invited to participate in unit operation and maintenance of management.Besides, operating room managers conducted regular sampling analysis of five micro-environmental indicators, including temperature, humidity, differential pressure, illumination and decibel, in 100-level, 1 000-level, 10 000-level, 100 000-level clean operating rooms, and compared the test results with parameters of target environmental indicators regulated in national standards.Then the test results of annual bacteria subsidence from different operating rooms in clean operating department were compared.Results Under the management of professionals, the values of temperature, humidity, differential pressure, illumination and decibel from micro-environment in 100-level, 1 000-level, 10 000-level, 100 000-level clean operating rooms were in accordance with rated values in the industry with

  4. Operational stability testing of biogas engine with varied fuel component%变组分生物质燃气发动机循环变动试验研究

    Institute of Scientific and Technical Information of China (English)

    陈雷; 宋鹏; 杜宝国; 隆武强; 曾文; 马洪安

    2016-01-01

    In view of the poor stability and large cyclic variations in biogas engines, experimental studies on variable component biomass ( CH4-H2-CO-CO2 ) engine operation stability were carried out on a 337 cc displacement spark ignition engine by employing bottled gases which includes CH4 , H2 and CO2 to simulate actual biogas.The influences of fuel and CO2 components on combustion stability were investigated, and an influence law of cyclic changes in the engine with changes in the fuel component was developed.The results show that engine operation stability is influenced strongly by fuel component.Under low H2 content conditions, CO2 content has a dominant effect on combustion stability.An increase in CO2 content could induce an increase in cyclic variation and lead to partial burning under special circumstances .However, under high H2 content conditions, the effect of CO2 content on combustion weakens gradually, and the effect is not obvious when the H2 content reaches 50%.Under constant H2 content conditions, variations in CH4 and CO contents have no obvious influence on biogas combustion.%针对生物质燃气发动机运转稳定性差、循环变动大的问题,采用瓶装气体模拟实际生物质燃气的方法,在一台火花点火发动机上进行了变组分生物质燃气(CH 4-H2-CO-CO2)发动机的循环变动试验研究,分析了各可燃组分及CO2含量变化对发动机工作稳定性的影响,给出了燃料组分变化对发动机循环变动的影响规律。结果表明,燃料组分变化对燃烧稳定性的影响情况有较大差异:在当量比固定且燃料组分中H2含量较少时,CO2含量对燃烧稳定性起支配作用,CO2含量的增加导致循环变动增加,在个别条件下导致部分燃烧的发生;随H2含量增加,CO2含量对燃烧的影响作用逐渐减弱,当H2含量达到50%时CO2含量对燃烧的影响作用已不明显;H2含量固定时,CH4及CO含量的变化对混合燃料的燃烧并无明显影响。

  5. 风电集中并网运行对系统电压稳定性的影响研究%Study on System Impacts of Voltage Stability about Wind Power Concentrated Grid-Connected Operation

    Institute of Scientific and Technical Information of China (English)

    付小伟; 王进; 郭伟; 郑剑武

    2012-01-01

    Uncertainty of wind power resources and operating characteristics of the wind turbine itself makes wind turbine output power of change, which is equivalent to big power system disturbances. With the gradual increase of wind power penetration, it's high time to study the Voltage stability of large grid-connected wind power operation problem, it detail analyses the effect of voltage stability about wind farm under different wind speed model, installed capacity of wind power and fault clearing time. The static synchronous compensator applied to wind farms, according to the simulation results, the voltage restore rapidly after the faults cleared, so the low voltage ride through(LVRT)capability of wind turbines is improved, ensures the continuous operation of wind turbines and the security of the power network%风力资源的不确定性和风电机组本身的运行特性使风电机组的输出功率随之变化,对电力系统来说相当于一个大的扰动。随着风电穿透功率的逐步增加,大规模风电并网运行的电压稳定性问题亟待研究。利用Matlab/Simulink建立风电场仿真模型,详细分析了不同风速模型、风电场装机容量及短路故障切除时间下风电场对电压稳定性的影响。把静止同步补偿器(StaticSynchronousCompensator,STATCOM)应用到风电场中,仿真结果表明,STATCOM可以改善风电场的电压稳定性,提高了风电机组的低电压穿越能力,确保风电机组连续运行及电网安全稳定。

  6. Stability Functions

    CERN Document Server

    Burns, Daniel; Wang, Zuoqin

    2008-01-01

    In this article we discuss the role of stability functions in geometric invariant theory and apply stability function techniques to problems in toric geometry. In particular we show how one can use these techniques to recover results of Burns-Guillemin-Uribe and Shiffman-Tate-Zelditch on asymptotic properties of sections of holomorphic line bundles over toric varieties.

  7. Comparison of the applicability of commercial computer programs to study the thermodynamic stability of metal oxides; Vertailu kaupallisten laskentaohjelmien soveltuvuudesta metallioksidien termodynaamiseen stabiilisuustarkasteluun

    Energy Technology Data Exchange (ETDEWEB)

    Koukkari, P.; Olin, M. [VTT Chemical Technology, Espoo (Finland); Laitinen, T. [VTT Manufacturing Technology, Espoo (Finland); Sippola, H. [GEM Systems Oy (Finland)

    1999-04-01

    The oxide films formed on primary circuit surfaces incorporate radioactive species from the coolant and influence different corrosion phenomena in nuclear power plants. One approach to improve the understanding on the formation and properties of these oxide films is to evaluate their thermodynamic stability. The aim of this work was to compare and demonstrate the applicability of different commercial thermodynamic computer programs to model the deposition of oxides containing iron, chromium and/or nickel in various nuclear power plant environments. The programs considered in this evaluation comprised the EQ3/6 program and a product group including such products as ChemSage, HSC (including SOLGASMIX and GIBBS), H+PLUS and ChemSheet. In the group the transfer of data between different products is relatively easy. The goal was to find out which programs can be applied to evaluate the stability of oxide films, but not to assess the absolute accuracy of the calculations. The evaluation was done by means of applying the programs to calculate the stability of pure and mixed oxides of iron, nickel and chromium on stainless steel both in WWER and BWR conditions at different temperatures and coolant compositions. The comparison showed that EQ3/6 is suitable for most thermodynamic calculations. EQ3/6 can be characterised as a professional tool, for which no commercial training is available. ChemSage is a versatile and reliable program, which can be well used together with HSC and H+PLUS. ChemSage is mainly a professional tool, while HSC is easier to operate by an occasional user. Commercial training and support is available for both ChemSage and HSC. ChemSheet has been designed to utilise the properties of ChemSage in a user-friendly spreadsheet environment. All the products tested calculate thermodynamic equilibrium. Thus they are suitable to characterise such conditions in which the formation of a certain phase is or becomes possible. On the other hand, the modelling of the

  8. Ensuring Stability

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    "Stable"will be a key word for China’s economy in 2012.That’s the beat set at the annual Central Economic Work Conference held in Beijing on December 12-14,which reviewed this year’s development and mapped out plans for the next year.Policymakers at the conference decided to keep macroeconomic policies stable,seek a stable and relatively fast economic growth,stabilize consumer prices and maintain social stability in 2012.On the basis of stability,the government will transform the development model,deepen reform and improve people’s livelihood.

  9. Army Medical Department Support to Stability Operations

    Science.gov (United States)

    2007-02-28

    25 million dead and another 40 million infected , continues to spread at an alarming rate and nervous eyes are turned toward an anticipated influenza...million men , women, and children suffer from chronic malnutrition that often comes in the aftermath of natural and man-made disasters.10 According to...included cholera, diphtheria, leishmaniasis, malaria, measles, pertussis, poliomyelitis, schistosomiasis , typhoid fever, and the chronic malnutrition

  10. Improving Capacity for Stabilization and Reconstruction Operations

    Science.gov (United States)

    2009-01-01

    Olga Oliker, Nora Bensahel, Derek Eaton, S. Jamie Gayton , Brooke Stearns Lawson, Jeffrey Martini, John L. Nasir, Michelle Parker, Jerry M. Sollinger... Gayton , Brooke Stearns Lawson, Jeffrey Martini, John L. Nasir, Michelle Parker, Jerry M. Sollinger, and Kayla M. Williams, Guidebook for Providing

  11. Structure, function and operational stability of peroxidases

    NARCIS (Netherlands)

    Haandel, van M.J.H.

    2000-01-01

    This PhD project was started in 1995 and was supported by the dutch Ministry of Economic Affairs through the programme "IOP Catalysis". The main goal of "IOP catalysis" is to obtain clean and more efficient technologies, to improve the quality of the Dutch fine chemistry. Biocatalysis provides a way

  12. Stability, Security, Transition and Reconstruction Operations Conference

    Science.gov (United States)

    2010-12-16

    at the U.S. Embassy in Bogota , Colombia from 2006-2009, managing the full range of U.S. bilateral relations in Colombia. He was an official...Assistance Program (VDAP) Examples of USG Interagency Collaboration Working with Civil Society • Zambezi River Basin Initiative

  13. Integrating Civilian Agencies in Stability Operations

    Science.gov (United States)

    2009-01-01

    because of the elements 1 Chris Huxham, Creating Collaborative Advantage, London: Sage, 1996. 2 Mancur Olson , The Logic of Collective Action: Public Goods...Nichols: Defense Reform for a New Strategic Era, Phase 2 Report, Washington, D.C.: Center for Strategic and International Studies, July 2005. Olson ... Mancur , The Logic of Collective Action: Public Goods and the Theory of Groups, Cambridge, MA: Harvard University Press, 1965. Bibliography 161

  14. Explosive Remnants of War in Stability Operations

    Science.gov (United States)

    2012-09-01

    Jute Kenaf Henequen Hemp Flax Hoopvine Elephant Grass Yucca (An agave) Water Reed Plantain Musamba Wood Fiber (kraft pulp) Maguey Lechuguilla Banana ...pulp) Maguey Lechuguilla Banana Leaf Guaney Figure 25. Natural Fibers Figure 26. Chopped fibers 1/2”–2” (1.5–5 cm) compared to raw fibers a...the weight and pressure of the FRBM against the I.D. form, it needs to be ridged. A ridged core wrapped in a thick plastic sheet is recommended. This

  15. The Doctrinal Basis for Medical Stability Operations

    Science.gov (United States)

    2010-01-01

    security forces (HNSFs). As most developing countries have poor health systems, all levels and branches of the health sector should be targeted including...actors will be available in more permissive environments, perhaps by orders of magnitude. Poor coordi- nation fragments efforts, weakens health systems...further services unless there is imme- diate threat to life, limb, or eyesight . MEDCAPs may have a role in rural areas without services but government

  16. Stabilizing Niger

    DEFF Research Database (Denmark)

    Hahonou, Eric Komlavi

    international intervention in Niger. Their main objective is to secure their own strategic, economic and political interests by strengthening the Nigerien authorities through direct intervention and capacity building activities. For western states reinforcing state security institutions and stabilizing elite...

  17. Major Combat Operations versus Stability Operations: Getting Army Priorities Correct

    Science.gov (United States)

    2011-05-19

    most likely encounter. Many defense analysts suggest that future conflict will be multimodal, combining various methods of warfare to increase both...fluid/maneuver http://usacac.leavenworth.amry.mil/BLOG/blogs/cac-t/archive/2009/04/09/semi-annual-ctc-conference-mco-and- comprehensive- aproach

  18. Classifier-ensemble incremental-learning procedure for nuclear transient identification at different operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Baraldi, Piero, E-mail: piero.baraldi@polimi.i [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Razavi-Far, Roozbeh [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Zio, Enrico [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Ecole Centrale Paris-Supelec, Paris (France)

    2011-04-15

    An important requirement for the practical implementation of empirical diagnostic systems is the capability of classifying transients in all plant operational conditions. The present paper proposes an approach based on an ensemble of classifiers for incrementally learning transients under different operational conditions. New classifiers are added to the ensemble where transients occurring in new operational conditions are not satisfactorily classified. The construction of the ensemble is made by bagging; the base classifier is a supervised Fuzzy C Means (FCM) classifier whose outcomes are combined by majority voting. The incremental learning procedure is applied to the identification of simulated transients in the feedwater system of a Boiling Water Reactor (BWR) under different reactor power levels.

  19. Space shuttle pogo studies. [systems stability

    Science.gov (United States)

    Coppolino, R. N.; Lock, M. H.; Rubin, S.

    1977-01-01

    Topics covered include: (1) pogo suppression for main propulsion subsystem operation; (2) application of quarter-scale low pressure oxidizer turbopump transfer functions; (3) pogo stability during orbital maneuvering subsystem operation; and (4) errors in frequency response measurements.

  20. Analysis by the Monte Carlo method of doses around the pool of storage of the control rods irradiated in a BWR reactor; Analisis mediante el metodo de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactror BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J.; Gallardo, S.

    2011-07-01

    The control rods of a boiling water reactor (BWR) are subject to a neutron flux and thus become activated during their stay in the reactor core. Activation occurs especially in the stainless steel components and impurities. The activity generated results in a dose around the bar, while it le unimportant in the reactor, but to be taken into account when removed f ron it. The bars drawn are stored on hangers placed in the storage pools of spent fuel f ron the plant. Each hanger 12 accommodates control rods and are arranged so that at least three meters of water abode the heads of the control rods. The dose received by potentially exposed workers who are in the vicinity of the storage must be calculated to ensure adequate protection of the came. This dose can be decreased significantly by changing the arrangement of the bars on hangers.

  1. 2015 Special Operations Essays

    Science.gov (United States)

    2015-05-01

    stability operations in Afghanistan.19 27 Goldstein: The First Women in SOF The persistent cultural stereotype of the female spy is of a hyper-sexual...outreach, and research in the science and art of joint special operations. JSOU provides education to the men and women of SOF and to those who...Human Domain in Warfare ................................ 11 The First Women in SOF: Women Operatives in the OSS and SOE as a Framework for the

  2. PARALLEL STABILIZATION

    Institute of Scientific and Technical Information of China (English)

    J.L.LIONS

    1999-01-01

    A new algorithm for the stabilization of (possibly turbulent, chaotic) distributed systems, governed by linear or non linear systems of equations is presented. The SPA (Stabilization Parallel Algorithm) is based on a systematic parallel decomposition of the problem (related to arbitrarily overlapping decomposition of domains) and on a penalty argument. SPA is presented here for the case of linear parabolic equations: with distrjbuted or boundary control. It extends to practically all linear and non linear evolution equations, as it will be presented in several other publications.

  3. ''Last experiences on ID BWR shroud inspection and the new developments to examine the below core plate areas''

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Willke, A.; Gonzalez, E.; Yague, L

    2001-07-01

    In recent years, the owners of BWR type nuclear power plants have had to address new inspection requirements relating to the core shroud inside the reactor vessel, the aim of which is to contain the fuel assemblies and provide support for the structures located in the upper part of the reactor. The shroud consists of a cylinder measuring some 40-50 mm in thickness, manufactured from various sections of AISI-304 stainless steel and INCONEL, joined by vertical and circumferential welds. The appearance of unstable cracks in these welds would directly affect the structural integrity of the component and the safety of the plant. As regards access to the core shroud and to the surface to be examined, two alternatives might be considered: inspection from outside the component, moving along the so-called annulus between the reactor vessel wall and the component (OD inspection), or from the interior (ID inspection). With a view to addressing this problem, Tecnatom has in recent years launched several projects, grouped under the generic name TEIDE, in order to develop scanners and NDT techniques achieving the maximum inspection coverage of this component. As regards inspection techniques, the decision was taken to carry out acquisition simultaneously using both ultrasonics (UT) and eddy currents (ET). (author)

  4. Statistical Safety Evaluation of BWR Turbine Trip Scenario Using Coupled Neutron Kinetics and Thermal Hydraulics Analysis Code SKETCH-INS/TRACE5.0

    Science.gov (United States)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.

  5. Macroeconomic stability

    DEFF Research Database (Denmark)

    Jespersen, Jesper

    2004-01-01

    It is demonstrated that full employment and sustainable development not necessarily are conflicting goals. On the other hand macroeconomic stability cannot be obtained without a deliberate labour sharing policy and a shift in the composition of private consumption away from traditional material...

  6. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Energy Technology Data Exchange (ETDEWEB)

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  7. Achieving Extreme Utilization of Excitons by an Efficient Sandwich-Type Emissive Layer Architecture for Reduced Efficiency Roll-Off and Improved Operational Stability in Organic Light-Emitting Diodes.

    Science.gov (United States)

    Wu, Zhongbin; Sun, Ning; Zhu, Liping; Sun, Hengda; Wang, Jiaxiu; Yang, Dezhi; Qiao, Xianfeng; Chen, Jiangshan; Alshehri, Saad M; Ahamad, Tansir; Ma, Dongge

    2016-02-10

    It has been demonstrated that the efficiency roll-off is generally caused by the accumulation of excitons or charge carriers, which is intimately related to the emissive layer (EML) architecture in organic light-emitting diodes (OLEDs). In this article, an efficient sandwich-type EML structure with a mixed-host EML sandwiched between two single-host EMLs was designed to eliminate this accumulation, thus simultaneously achieving high efficiency, low efficiency roll-off and good operational stability in the resulting OLEDs. The devices show excellent electroluminescence performances, realizing a maximum external quantum efficiency (EQE) of 24.6% with a maximum power efficiency of 105.6 lm W(-1) and a maximum current efficiency of 93.5 cd A(-1). At the high brightness of 5,000 cd m(-2), they still remain as high as 23.3%, 71.1 lm W(-1), and 88.3 cd A(-1), respectively. And, the device lifetime is up to 2000 h at initial luminance of 1000 cd m(-2), which is significantly higher than that of compared devices with conventional EML structures. The improvement mechanism is systematically studied by the dependence of the exciton distribution in EML and the exciton quenching processes. It can be seen that the utilization of the efficient sandwich-type EML broadens the recombination zone width, thus greatly reducing the exciton quenching and increasing the probability of the exciton recombination. It is believed that the design concept provides a new avenue for us to achieve high-performance OLEDs.

  8. 计算机指点装置的操作可控性及稳定性探讨%Discussion on Computer Pointing Devices Operating Controllability and Stability

    Institute of Scientific and Technical Information of China (English)

    宋铁石; 李同伟; 王冠

    2014-01-01

    在当前计算机技术中,计算机指点装置是极为重要的组成部分,它是图形用户界面中常用到的技术,目前较为常见的指点装置就是鼠标。但如果计算机使用者长期使用鼠标,会导致其出现肌肉骨骼不适等问题。因此,必须要积极研究指点装置的操作特点,优化指点装置的设计,发挥其最大的作用,减少其在使用过程中对人体的伤害。稳定性以及可控性是评价计算机指点装置性能的主要指标,针对此指标做了详细的探讨。%In the current computer technology, computer pointing device is a very important part, it is commonly used in the graphical user interface to the technology, the more common is the mouse pointing device. But if the computer users uses the mouse for the long time,leading to problems of musculoskeletal discomfort. Therefore, the operating characteristics must actively study the pointing device, design optimization of pointing device, produce its biggest effect, reduce the harm to human body in the use process. Stability and controllability is the main index to evaluate the performance of computer pointing device, this paper discusses briefly.

  9. Seismic safety margins research program. Phase I. Project VII: systems analysis specifications of computational approach. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Collins, J.D.; Hudson, J.M.; Chrostowski, J.D.

    1979-02-01

    A computational methodology is presented for the prediction of core melt probabilities in a nuclear power plant due to earthquake events. The proposed model has four modules: seismic hazard, structural dynamic (including soil-structure interaction), component failure and core melt sequence. The proposed modules would operate in series and would not have to be operated at the same time. The basic statistical approach uses a Monte Carlo simulation to treat random and systematic error but alternate statistical approaches are permitted by the program design.

  10. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2016-04-15

    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

  11. 跨临界CO2引射制冷循环运行稳定性研究%The Operation Stability Research on the Transcritical CO2 Ejector Expansion Refrigeration Cycle

    Institute of Scientific and Technical Information of China (English)

    邓建强

    2011-01-01

    Present work analyzed the operation stability of the transcritical CO2 ejector expansion refrigeration cycle. Thermodynamic models of two kinds adjusting schemes which maintain cycle stable, i.e. feedback vapor and two-stage evaporation were set up. The simulation result shows that cycle with vapor feedback adjusting control has highly similar refrigeration effect with the basic ejector expansion cycle which controlled by the entrainment ratio. The cycle with two-stage evaporation improves the refrigeration performance obviously. The adjustable district of vapor feedback control becomes narrower with the entrainment ratio decreasing. By contrast, the adjustable district of the two-stage evaporation control broadens. With regards to the vapor feedback adjusting control, considering the optimize of the COP and the wider adjustable working condition, the ejector coefficient is suitable in 0.6-0.7.%分析了跨临界CO2引射制冷循环的运行稳定性问题,对蒸气反馈和两级蒸发两种致引射系统稳定的调控方案构建了热力学模型。模拟研究表明蒸气反馈调控系统具有与基本引射循环调控引射系数致系统稳定非常相似的制冷效果。而两级蒸发调控可明显提高系统的制冷性能。引射系数越小,蒸气反馈调控的可调工况区间越窄,两级蒸发方案的可调工况区间越宽。对于蒸气反馈调控,同时考虑COP优化和具较宽的可调工况,引射系数在0.6-0.7间较合适。

  12. Analysis of non-fault tripping criterion operation for area-based stability control device%区域型安全稳定控制装置无故障跳闸判据运行分析

    Institute of Scientific and Technical Information of China (English)

    王伟; 陈军; 余锐; 朱小红

    2012-01-01

    系统分析了区域型安控装置判断线路、主变等元件的无故障跳闸判据,针对电网潮流变化、直流输电系统双极闭锁、直流输电减负荷等运行情况下该判据暴露的问题,指出安控装置仅靠本地电气量判断线路本侧及对侧的无故障跳闸存在着一定的局限性,尤其是在直流输电的近端电网.根据安控装置在电网的各种运行工况下发生无故障跳闸后的控制要求,对直流输电近端电网的安控装置增加了开关量信号作为无故障跳闸的主判据,电气量作为辅助判据,并结合实际工程完成了无故障跳闸判据整改,较好地解决了该判据不完善的问题.%This paper analyzes the non-fault tripping criterion for components of area-based control device such as transmission line, generator transformer and so on. In view of the problem that this criterion exposes in the situation of grid power flow change, HVDC system double pole locking, and HVDC load reduction, there are limitations to identify the non-fault line tripping at one side and the opposite side only through local electrical measurements, especially in the near-end grid of HVDC. According to the control requirements of control device after non-fault line tripping occurring in different operation conditions, switching value signals and electrical quantity are added to the control device of HVDC near-end grid to be the main criterion and the auxiliary criterion respectively. A new criterion of non-fault tripping is proposed with the rectification of security and stability control device of Sichuan grid, which solves the problem of the existing criterion.

  13. Experimental data report for test TS-4, Reactivity initiated accident test in the NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1994-01-01

    本報告書は1991年1月に実施した照射済BWR燃料を用いた4回目の反応度事故模擬実験であるTS-4について、実験データをまとめたものである。TS-4実験に使用した試験燃料は、初期濃縮度2.79%であり、日本原子力発電(株)の敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した燃料の燃焼度は26GWd/tUであった。NSRRにおける照射実験は、BWRのコールドスタートアップ条件を模擬した大気圧・室温の静止水冷却条件下で行い、公称発熱量は110pm5cal/g・fuel(ピークエンタルピ89pm4cal/g・fuel)を与えた。その結果、燃料破損は生じなかった。実験条件、実験方法、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  14. Experimental data report for test TS-3; Reactivity initiated accident test in the NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1993-01-01

    本報告書は、1990年9月に実施した照射済BWR燃料を用いた3回目の反応度事故模擬実験であるTS-3について実験データをまとめたものである。TS-3実験に使用した試験燃料は初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3GWd/tUであった。NSRRにおける照射実験は、大気圧・室温の静止水冷却条件下で行い、発熱量は94pm4cal/g・fuel(ピークエンタルピ88pm4cal/g・fuel)を与えた。その結果燃料破損は生じなかった。実験条件、実験方法、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  15. Kohonen mapping of the crack growth under fatigue loading conditions of stainless steels in BWR environments and of nickel alloys in PWR environments

    Science.gov (United States)

    Urquidi-Macdonald, Mirna

    2008-09-01

    In this study, crack growth rate data under fatigue loading conditions generated by Argonne National Laboratories and published in 2006 were analyzed [O.K. Chopra, B. Alexandreanu, E.E. Gruber, R.S. Daum, W.J. Shack, Argonne National Laboratory, NUREG CR 6891-series ANL 04/20, Crack Growth Rates of Austenitic Stainless Steel Weld Heat Affected Zone in BWR Environments, January, 2006; B. Alexandreanu, O.K. Chopra, H.M. Chung, E.E. Gruber, W.K. Soppet, R.W. Strain, W.J. Shack, Environmentally Assisted Cracking in Light Water Reactors, vol. 34 in the NUREG/CR-4667 series annual report of Argonne National Laboratory program studies for Calendar (Annual Report 2003). Manuscript Completed: May 2005, Date Published: May 2006], and reported by DoE [B. Alexandreanu, O.K. Chopra, W.J. Shack, S. Crane, H.J. Gonzalez, NRC, Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments, NUREG/CR-6964, May 2008]. The data collected were measured on austenitic stainless steels in BWR (boiling water reactor) environments and on nickel alloys in PWR (pressurized water reactor) environments. The data collected contained information on material composition, temperature, conductivity of the environment, oxygen concentration, irradiated sample information, weld information, electrochemical potential, load ratio, rise time, hydrogen concentration, hold time, down time, maximum stress intensity factor ( Kmax), stress intensity range (Δ Kmax), crack length, and crack growth rates (CGR). Each position on that Kohonen map is called a cell. A Kohonen map clusters vectors of information by 'similarities.' Vectors of information were formed using the metal composition, followed by the environmental conditions used in each experiments, and finally followed by the crack growth rate (CGR) measured when a sample of pre-cracked metal is set in an environment and the sample is cyclically loaded. Accordingly

  16. Hydraulic modeling and simulation of a System Division of Essential Service Water in a BWR plant with Flow master; Modelo hidraulico y simulacion de una division del Sistema de Agua de Servicio Esencial de una central BWR con Flowmaster

    Energy Technology Data Exchange (ETDEWEB)

    Vegazo Juzgado, L.; Rodriguez Garcia, G. M.; Mota Coloma, M.

    2012-07-01

    At the conclusion of the project can say that Flow master is a simulation tool that allows you to create your model from a library of components and obtain useful results from the point of view of the operation, engineering and maintenance. Compared to previous software from the point of view of use, can comment that Flow master is a tool which has an intuitive and user-friendly interaction between the user and the program thus facilitating the modeling of the system and definition of the components of same.

  17. 46 CFR 154.1809 - Loading and stability manual.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Loading and stability manual. 154.1809 Section 154.1809... stability manual. (a) No person may operate a vessel unless that vessel has on board a loading and stability manual. (b) The loading and stability manual must contain: (1) Information that enables the master...

  18. Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands` PINK programme (Joint project of ECN, IRI, KEMA and GKN)

    Energy Technology Data Exchange (ETDEWEB)

    Gruppelaar, H. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Klippel, H.T. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Verhagen, F.C.M. [Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands); Bruggink, J.C. [Gemeenschappelijke Kernenergiecentrale Nederland N.V., Dodewaard (Netherlands)

    1993-11-01

    Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

  19. Stabilizing brokerage.

    Science.gov (United States)

    Stovel, Katherine; Golub, Benjamin; Milgrom, Eva M Meyersson

    2011-12-27

    A variety of social and economic arrangements exist to facilitate the exchange of goods, services, and information over gaps in social structure. Each of these arrangements bears some relationship to the idea of brokerage, but this brokerage is rarely like the pure and formal economic intermediation seen in some modern markets. Indeed, for reasons illuminated by existing sociological and economic models, brokerage is a fragile relationship. In this paper, we review the causes of instability in brokerage and identify three social mechanisms that can stabilize fragile brokerage relationships: social isolation, broker capture, and organizational grafting. Each of these mechanisms rests on the emergence or existence of supporting institutions. We suggest that organizational grafting may be the most stable and effective resolution to the tensions inherent in brokerage, but it is also the most institutionally demanding.

  20. Active Stabilization of Aeromechanical Systems

    Science.gov (United States)

    1993-01-05

    feedback, the resulting two-mode proportonal control law further extended the stable range of operation of the compresor . Stall flow coefficient was...active control of surge and stall in gas turbine engines. The use of small amplitude waves predicted by theory as stall precursors were tested with...stabilization system for rotating stall which was tested on both a single-stage and a three-stage axial compressor, increasing the stable operating range

  1. Sensitivity study on heuristic rules applied to the neutronic optimization of cells for BWR; Estudio de sensibilidad sobre reglas heuristicas aplicadas a la optimizacion neutronica de celdas para BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, J.; Martin del Campo M, C.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2004-07-01

    The objective of this work is to verify the validity of the heuristic rules that have been applied in the processes of radial optimization of fuel cells. It was examined the rule with respect to the accommodation of fuel in the corners of the cell and it became special attention on the influence of the position and concentration of those pellets with gadolinium in the reactivity of the cell and the safety parameters. The evaluation behaved on designed cells violating the heuristic rules. For both cases the cells were analyzed between infinite using the HELIOS code. Additionally, for the second case, it was behaved a stage more exhaustive where it was used one of the studied cells that it completed those safety parameters and of reactivity to generate the design of an assemble that was used to calculate with CM-PRESTO the behavior of the nucleus during three operation cycles. (Author)

  2. Improving the action requirements of technical specifications: A risk-comparison of continued operation and plant shutdown

    Energy Technology Data Exchange (ETDEWEB)

    Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States); Mankamo, T.

    1995-04-01

    When the systems needed to remove decay heat are inoperable or degraded, the risk of shutting down the plant may be comparable to, or even higher than, that of continuing power operation with the equipment inoperable while giving priority to repairs. This concern arises because the plant may not have sufficient capability for removing decay heat during the shutdown. However, Technical Specifications (TSs) often require {open_quotes}immediate{close_quotes} shutdown of the plant. In this paper, we present risk-based analyses of the various operational policy alternatives available in such situations, with an example application to the standby service water (SSW) system of a BWR. These analyses can be used to define risk-effective requirements for those standby safety systems under discussion.

  3. Effect of RCIC Operating Conditions on the Accident Scenario in Fukushima Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Park, Jong Hwa; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This study was conducted by using MELCOR 1.8.6. Fukushima unit 2 accident was analyzed using MELCOR in this study, and best estimate scenario with considering RCIC operating conditions was presented. Researches on the boiling water reactor (BWR) plant with reactor core isolation cooling (RCIC) system have been conducted. Research on the RCIC operation in Fukushima unit 2 was also conducted by Sandia National Laboratory. MELCOR analysis of the Fukushima unit 2 accident was conducted in the report and energy balance in wetwell was described by considering RCIC operation. However, the effect of RCIC operation condition on the accident scenario has not been studied. The operating conditions of RCIC system affect the pressures in wetwell and drywell, and the high pressure can make leakage path of fission product from PCV to reactor building. Thus it can be directly related with the amount of fission product which released to environment. In this study, severe accident on Fukushima unit 2 was analyzed considering the operating condition of RCIC system, and best estimated scenario was presented. In addition, the effect of RCIC turbine efficiency on the accident progression was examined. Energy balance in suppression chamber was also considered with discussion on the effect of torus room flooding level. It was found that the operating condition of RCIC turbine not only affects the variation of drywell pressure but also the amount of released fission products to environment. It was also confirmed that the RCIC turbine efficiency in the accident would be less than normal operating condition.

  4. Reframing Stability Operations: using Social Science to Identify Pillars of Stability Operations to Bridge the Gap Between the Principles of Joint Operations and Stability Operations Framework

    Science.gov (United States)

    2009-12-11

    cannot fully restore a person’s range of choices had they not engaged in detrimental behavior in the first place. Teen pregnancy and child...cause Zimbabwe, a country that used to produce vast excesses of food for distribution throughout Africa , to import food to keep supplies at minimum...the worst in  Africa  ‐ 24% in 2000. Adaptation Economic System of  Needs‐satisfier  production Barre gov’t stifled economic growth.   Insecurity during

  5. Experimental data report for test TS-5; Reactivity initiated accident test in the NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1995-01-01

    本報告書は、1993年度1月に実施した照射済BWR燃料を用いた5回目の反応度事故模擬実験であるTS-5について、実験データをまとめたものである。TS-4実験に使用した試験燃料は、初期濃縮度2.79%であり、日本原子力発電(株)の敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した燃料の燃焼度は26GWd/tUであった。NSRRにおける照射実験は、BWRのコールドスタートアップ条件を模擬した大気圧・室温の静止水冷却条件下で行い、公称発熱量は117pm5cal/g・fuel(ピークエンタルピ98pm4cal/g・fuel)を与えた。その結果燃料破損は生じなかった。なお、この実験では集合体中の燃料/水比を模した流路管中で燃料のパルス照射を行った。実験条件、実験方法、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  6. Nonlinear physical systems spectral analysis, stability and bifurcations

    CERN Document Server

    Kirillov, Oleg N

    2013-01-01

    Bringing together 18 chapters written by leading experts in dynamical systems, operator theory, partial differential equations, and solid and fluid mechanics, this book presents state-of-the-art approaches to a wide spectrum of new and challenging stability problems.Nonlinear Physical Systems: Spectral Analysis, Stability and Bifurcations focuses on problems of spectral analysis, stability and bifurcations arising in the nonlinear partial differential equations of modern physics. Bifurcations and stability of solitary waves, geometrical optics stability analysis in hydro- and magnetohydrodynam

  7. Can the UN Stabilize Mali? Towards a UN Stabilization Doctrine?

    Directory of Open Access Journals (Sweden)

    Arthur Boutellis

    2015-06-01

    Full Text Available Almost two years after the deployment of the United Nations Multidimensional Integrated Stabilization Mission in Mali (MINUSMA in July 2013, the increasing number of asymmetric terrorist attacks targeting UN peacekeepers – in the context of a drawn-out peace process – has raised a number of questions in Mali, the sub-region, and in New York, over the relevance and adequacies of MINUSMA’s mandate and capabilities. It also raises a broader issue, of whether the consent-based UN peacekeeping tool is appropriate and can be effective in carrying out stabilization mandates in such a context and what doctrine such operations should be based on. The UN is indeed under increasing pressure from host countries and some African troop-contributing countries to go on the offensive. Member States have also increasingly recognized terrorism and organized crime as a strategic threat, and while opposed to the UN directly engaging in counterterrorism (CT operations, some may wish to see the UN playing a greater stabilization role following the January 2013 French military intervention in Mali. However, little guidance and means have been given so far to UN missions for dealing with such threats and implementing effective stabilization mandates. The High-Level Panel on Peace Operations, which recently released its report, noted that the usage of the term “stabilization” by the UN requires clarification. This article analyses the complex and evolving nature of threats in northern Mali and implications for MINUSMA and describes the military and political tools – including mediation – so far available within and outside the UN. The article concludes that the UN is bound to move towards stabilization when and if deployed in contexts such as Mali’s if it wants to remain relevant. However, such a move should be based on an overarching UN stabilization doctrine and context-specific UN-wide stabilization strategies which are first and foremost political

  8. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  9. Operational amplifiers

    CERN Document Server

    Dostal, Jiri

    1993-01-01

    This book provides the reader with the practical knowledge necessary to select and use operational amplifier devices. It presents an extensive treatment of applications and a practically oriented, unified theory of operational circuits.Provides the reader with practical knowledge necessary to select and use operational amplifier devices. Presents an extensive treatment of applications and a practically oriented, unified theory of operational circuits

  10. Foaming volume and foam stability

    Science.gov (United States)

    Ross, Sydney

    1947-01-01

    A method of measuring foaming volume is described and investigated to establish the critical factors in its operation. Data on foaming volumes and foam stabilities are given for a series of hydrocarbons and for a range of concentrations of aqueous ethylene-glycol solutions. It is shown that the amount of foam formed depends on the machinery of its production as well as on properties of the liquid, whereas the stability of the foam produced, within specified mechanical limitations, is primarily a function of the liquid.

  11. Laguna Verde BWRs operational experience: steady-state fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G. F.; Bravo S, J. M. [Global Nuclear Fuel - Americas, 3901 Castle Hayne Road, Wilmington, 28401 North Carolina (United States); Casillas, J. L., E-mail: gabriel.cuevas-vivas@gnf.co [General Electric Hitachi Nuclear Energy, 1989 Little Orchard St. Romm 239, San Jose, 95125 California (United States)

    2010-10-15

    The two BWR at Laguna Verde nuclear power station are finishing 21 and 15 years of continuous successful operation as of 2010. During Unit 1 and 2 commercial operations only Ge/GNF fuel designs have been employed; fuel lattice designs 8 x 8 and 10 x 10 were used at the reactor, with an original licensed thermal power (OLTP: 1931 MWt) and the reactor's first power up-rates of 5%. GNF fuel will be also used for the second EPU to reach 120% of OLTP in the near future. Thermal and gamma traversing in-core probes (Tip) are used for power monitoring purposes along with the Ge (now GNF-A) core monitoring system, 3-dimensional Monicore{sup TM}. GNF-A has also participated by preparing the core management plan that is regularly fine-tuned in collaboration with Comision Federal de Electricidad (CFE owner of the Laguna Verde reactors). For determination of thermal margins and eigenvalue prediction, GNF-A employs the NRC-licensed steady-state core simulator PANAC11. Tip comparisons are routinely used to adapt power distributions for a better thermal margin calculation. Over the years, several challenges have appeared in the near and long term fuel management planning such as increasing cycle length, optimization of the thermal margins, rated power increase, etc. Each challenge has been successfully overcome via operational strategy, code improvements and better fuel designs. This paper summarizes Laguna Verde Unit 1 and 2 steady-state performance from initial commercial operation, with a discussion of the nuclear and thermal-hydraulic design features, as well as of the operational strategies that set and interesting benchmark for future fuel applications, code development and operation of the BWRs. (Author)

  12. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  13. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  14. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator; Modelado de la dinamica de la vasija y circuitos de recirculacion de una nucleoelectrica tipo BWR como parte del simulador universitario SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R.A. [DEPFI, Campus Morelos, en IMTA, Jiutepec, Morelos (Mexico)]. e-mail: rsanchez_15@yahoo.com.mx

    2003-07-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  15. Matlab Stability and Control Toolbox: Trim and Static Stability Module

    Science.gov (United States)

    Crespo, Luis G.; Kenny, Sean P.

    2006-01-01

    This paper presents the technical background of the Trim and Static module of the Matlab Stability and Control Toolbox. This module performs a low-fidelity stability and control assessment of an aircraft model for a set of flight critical conditions. This is attained by determining if the control authority available for trim is sufficient and if the static stability characteristics are adequate. These conditions can be selected from a prescribed set or can be specified to meet particular requirements. The prescribed set of conditions includes horizontal flight, take-off rotation, landing flare, steady roll, steady turn and pull-up/ push-over flight, for which several operating conditions can be specified. A mathematical model was developed allowing for six-dimensional trim, adjustable inertial properties, asymmetric vehicle layouts, arbitrary number of engines, multi-axial thrust vectoring, engine(s)-out conditions, crosswind and gyroscopic effects.

  16. CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR usando los codigos OpenFOAM y GasFlow

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.

    2014-07-01

    using a limited number of semi-empirical data, and instead, mathematical relationships are used taking into account the various physical phenomena as well the interactions that occur among them, such as heat transfer between the fluid and the solid walls condensation of water vapor on the walls, the turbulent effects in areas of restricted passage, etc. Taking into account these advantages, this study presents a qualitative and quantitative comparison between the CFD codes OpenFOAM and Gas-Flow related to the transport phenomena of Hydrogen and other gases in the primary containment of a BWR reactor. Gas-Flow is a code of commercial license that is well validated, developed in Germany to analyze the transport of gases in nuclear reactor containments. On the other hand, OpenFOAM is an open source CFD code offering several solvers for different phenomena assessments, in this work, the reacting Foam solver is used because it has a strong similarity to the intended application of Hydrogen transport. In this thesis the results obtained using the reacting Foam solver of OpenFOAM for the calculation of transport of Hydrogen are compared with the results of the Gas-Flow code in order to assess if it is feasible to use the open source code OpenFOAM in the case of Hydrogen transport in primary containment of a BWR reactor. Some differences in the qualitative and quantitative results from both codes were found, the differences (with a maximum error rate of 4%) in the quantitative results were found are small and are considered more than acceptable for this type of analysis, moreover, these differences are mainly attributed to the transport models used, mainly because OpenFOAM uses a homogeneous mixture model and Gas-Flow a heterogeneous one. Implementing appropriate solvers in codes like OpenFOAM has the goal to develop own tools that are applicable to the transport of Hydrogen in the primary containment of a BWR reactor and thus, to gain some independence while not relying on

  17. Stability of Boolean Multiplex Networks

    CERN Document Server

    Cozzo, Emanuele; Moreno, Yamir

    2012-01-01

    We extend the formalism of Random Boolean Networks with canalizing rules to multilevel complex networks. The formalism allows to model genetic networks in which each gene might take part in more than one signaling pathway. We use a semi-annealed approach to study the stability of this class of models when coupled in a multiplex network and show that the analytical results are in good agreement with numerical simulations. Our main finding is that the multiplex structure provides a mechanism for the stabilization of the system and of chaotic regimes of individual layers. Our results help understanding why some genetic networks that are theoretically expected to operate in the chaotic regime can actually display dynamical stability.

  18. Plutonium inventories for stabilization and stabilized materials

    Energy Technology Data Exchange (ETDEWEB)

    Williams, A.K.

    1996-05-01

    The objective of the breakout session was to identify characteristics of materials containing plutonium, the need to stabilize these materials for storage, and plans to accomplish the stabilization activities. All current stabilization activities are driven by the Defense Nuclear Facilities Safety Board Recommendation 94-1 (May 26, 1994) and by the recently completed Plutonium ES&H Vulnerability Assessment (DOE-EH-0415). The Implementation Plan for accomplishing stabilization of plutonium-bearing residues in response to the Recommendation and the Assessment was published by DOE on February 28, 1995. This Implementation Plan (IP) commits to stabilizing problem materials within 3 years, and stabilizing all other materials within 8 years. The IP identifies approximately 20 metric tons of plutonium requiring stabilization and/or repackaging. A further breakdown shows this material to consist of 8.5 metric tons of plutonium metal and alloys, 5.5 metric tons of plutonium as oxide, and 6 metric tons of plutonium as residues. Stabilization of the metal and oxide categories containing greater than 50 weight percent plutonium is covered by DOE Standard {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides{close_quotes} December, 1994 (DOE-STD-3013-94). This standard establishes criteria for safe storage of stabilized plutonium metals and oxides for up to 50 years. Each of the DOE sites and contractors with large plutonium inventories has either started or is preparing to start stabilization activities to meet these criteria.

  19. Quantum Contextuality with Stabilizer States

    Directory of Open Access Journals (Sweden)

    Jiri Vala

    2013-06-01

    Full Text Available The Pauli groups are ubiquitous in quantum information theory because of their usefulness in describing quantum states and operations and their readily understood symmetry properties. In addition, the most well-understood quantum error correcting codes—stabilizer codes—are built using Pauli operators. The eigenstates of these operators—stabilizer states—display a structure (e.g., mutual orthogonality relationships that has made them useful in examples of multi-qubit non-locality and contextuality. Here, we apply the graph-theoretical contextuality formalism of Cabello, Severini and Winter to sets of stabilizer states, with particular attention to the effect of generalizing two-level qubit systems to odd prime d-level qudit systems. While state-independent contextuality using two-qubit states does not generalize to qudits, we show explicitly how state-dependent contextuality associated with a Bell inequality does generalize. Along the way we note various structural properties of stabilizer states, with respect to their orthogonality relationships, which may be of independent interest.

  20. Lead exposure reduction in workers using stabilizers in PVC manufacture: effects of a new encapsulated stabilizer

    Energy Technology Data Exchange (ETDEWEB)

    Fischbein, A. (Mount Sinai Medical Center, New York, NY); Thornton, J.C.; Berube, L.; Villa, F.; Selikoff, I.J.

    1982-09-01

    The handling of lead-containing powdered stabilizers may be associated with excessive lead exposure and occupational lead poisoning. An encapsulated stabilizer has been designed, which is shown to result in a marked reduction in air lead levels and a concommitant decrease in blood lead concentrations in a group of blender operators manufacturing PVC electrical cable coating.