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Sample records for study bwr instabilities

  1. Experimental study on thermo-hydraulic instability on reduced-moderation natural circulation BWR concept

    International Nuclear Information System (INIS)

    Watanabe, Noriyuki; Subki, M.H.; Kikura, Hiroshige; Aritomi, Masanori

    2003-01-01

    Reduced-moderation natural circulation BWR has been promoted to solve the recent challenges in BWR nuclear power technology problems as one of advanced small and medium-sized reactors equipped with the passive safety features in conformity with the natural law. However, the elimination of recirculation pumps and a high-density core due to the increase of conversion ratio could cause various thermo-hydraulic instabilities especially during the start-up stage. The occurrences of the thermo-hydraulic instabilities are not desirable and it is one of the main challenges in establishing reduced-moderation natural circulation BWR as a commercial reactor. The purpose of this present study is to experimentally investigate the driving mechanism of the thermo-hydraulic instabilities and the effect of system pressure on the unstable flow patterns. Hence, as the fundamental research for this study, a natural circulation loop that carries boiling fluid with parallel boiling channel has been constructed. Channel gap that has been set at 2 mm in order to simulate reduced-moderation reactor core. Pressure ranges of 0.1 up to 0.7 MPa, input heat flux range of 0 ou to 577 kW/m 2 , and inlet subcooling temperatures of 5, 10, and 15 K respectively, are imposed in the experiments. This experiment clarifies that changes in unstable flow patterns with increase in heat flux can be classified into two in response to system pressure range. In case of atmospheric pressure, unstable flow patters has been classified in beyond order, (1) in-phase geysering, (2) transition oscillation combined with both features of in-phase geysering and natural circulation oscillation, (3) natural circulation oscillation induced by hydrostatic head fluctuation, (4) density wave oscillation, and finally (5) stable boiling two-phase flow. On the other hand, in the system pressure range from 0.2 to 0.7 MPa, unstable patters have been dramatically changed in the following order (1) out-of-phase geysering, (2

  2. Water level measurement uncertainty during BWR instability

    International Nuclear Information System (INIS)

    Torok, R.C.; Derbidge, T.C.; Healzer, J.M.

    1994-01-01

    This paper addresses the performance of the water-level measurement system in a boiling water reactor (BWR) during severe instability oscillations which, under some circumstances, can occur during an anticipated transient without SCRAM (ATWS). Test data from a prototypical mock-up of the water-level measurement system was used to refine and calibrate a water-level measurement system model. The model was then used to predict level measurement system response, using as boundary conditions vessel pressures calculated by ppercase RETRAN for an ATWS/instability event.The results of the study indicate that rapid pressure changes in the reactor pressure vessel which cause oscillations in downcomer water level, coupled with differences in instrument line lengths, can produce errors in the sensed water level. Using nominal parameters for the measurement system components, a severe instability transient which produced a 0.2 m peak-to-minimum water-level oscillation in the vessel downcomer was predicted to produce pressure difference equivalent to a 0.7 m level oscillation at the input to the differential pressure transmitter, 0.5 m oscillation at the output of the transmitter, and an oscillation of 0.3 m on the water-level indicator in the control room. The level measurement system error, caused by downcomer water-level oscillations and instrument line length differential, is mitigated by damping both in the differential pressure transmitter used to infer level and in the control room display instrument. ((orig.))

  3. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  4. BWR regional instability model and verification on ringhals-1 test

    International Nuclear Information System (INIS)

    Hotta, Akitoshi; Suzawa, Yojiro

    1996-01-01

    Regional instability is known as one type of the coupled neutronic-thermohydraulic phenomena of boiling water reactors (BWRs), where the thermohydraulic density wave propagation mechanism is predominant. Historically, it has been simulated by the three-dimensional time domain code in spite of its significant computing time. On the other hand, there have been proposals to apply the frequency domain models in regional instability considering the subcriticality of the higher neutronic mode. However, their application still remains in corewide instability mainly because of the lack of more detailed methodological and empirical studies. In this study, the current version of the frequency domain model was extended and verified based on actual core regional instability measurement data. The mathematical model LAPUR, the well-known frequency domain stability code, was reviewed from the standpoint of pure thermohydraulics and neutronic-thermohydraulic interaction mechanisms. Based on the ex-core loop test data, the original LAPUR mixed friction and local pressure loss model was modified, taking into account the different dynamic behavior of these two pressure-loss mechanisms. The perturbation term of the two-phase friction multiplier, which is the sum of the derivative of void fraction and subcool enthalpy, was adjusted theoretically. The adequacy of the instability evaluation system was verified based on the Ringhals unit 1 test data, which were supplied to participants of the Organization for Economic Cooperation and Development/Nuclear Energy Agency BWR Stability Benchmark Project

  5. An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Masashi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-07-01

    This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

  6. 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.

  7. Prony's method application for BWR instabilities characterization

    International Nuclear Information System (INIS)

    Castillo, Rogelio; Ramírez, J. Ramón; Alonso, Gustavo; Ortiz-Villafuerte, Javier

    2015-01-01

    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

  8. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  9. A New Method for Early Anomaly Detection of BWR Instabilities

    International Nuclear Information System (INIS)

    Ivanov, K.N.

    2005-01-01

    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.

  10. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

    1990-03-01

    Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

  11. Noise analysis of Forsmark 1 data to investigate BWR core local instability

    International Nuclear Information System (INIS)

    Oguma, R.

    1998-04-01

    BWR core local instability was experienced at Forsmark 1 (F1) during reactor operation in cycle 16. The event has been studied by applying noise analysis and stability calculations to get insight into the event as well as to identify the cause of local instability. The present report is concerned with noise analysis of data collected during start-up in cycle 17. The results of the current study indicates: The F1 core is quite stable in cycle 17. The max. decay ratio (DR) value of 0.37 was obtained from the stability evaluation of an APRM (average power range monitor) and LPRM (local power range monitor) signals measured at 66% (APRM) of reactor power and 4252 Kg/s (SA-HC) of core flow. Compared with the power profile in cycle 17 (as well as in reactor F2), the core in cycle 16 had an extreme power profile with high power and bottom-shifted axial peak in the core periphery esp. at the four quadrant corners. Such a profile decreases the stability margin in the region. It is a common observation that the DR obtained from APRM tends to be higher than that from LPRM if the global instability mechanism is dominant in the core, and vice versa. The comparison of global and local DR values should be an effective method for detecting local instability during the reactor operation. In order to detect the local instability it is important to evaluate the core stability with sufficient number of LPRMs so as to cover the whole core cross section together with APRMs

  12. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Alessandro Petruzzi; Shin Chin; Kostadin Ivanov; Asok Ray; Fan-Bill Cheung

    2005-01-01

    Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order 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 model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

  13. Analysis of BWR out-of-phase instabilities in the frequency domain

    International Nuclear Information System (INIS)

    Farawila, Y.M.; Pruitt, D.W.; Kreuter, D.

    1992-01-01

    During startup or because of an inadvertent recirculation pump trip, a boiling water reactor (BWR) may operate at relatively low flow and high power conditions. At these conditions, a BWR is susceptible to coupled flow and power oscillations that could result in undesirable reactor scram unless appropriate countermeasures are taken. This contribution to analytical methods has been developed to address in part a general industrywide and regulatory concern about BWR stability initiated by the LaSalle 2 instability event in March 1988. This work is designed to extend the capability of the one-dimensional parallel channel frequency domain code STAIF to predict the regional oscillation decay ratio. The basic theory follows that developed by March-Leuba and Blakeman, where the oscillation mechanism is identified as the excitation of a subcritical neutronic mode with a constant core pressure drop boundary condition. The improvements to the basic theory include applying the theory to one-dimensional neutronics instead of point kinetics and taking account of the actual three-dimensional harmonic flux distribution

  14. Effect of the inlet throttling on the thermal-hydraulic instability of the natural circulation BWR

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Inada, Fumio; Yoneda, Kimitoshi

    1997-01-01

    Although it is well-established that inlet restriction has a stabilizing for forced circulation BWR, the effect of inlet on the thermal-hydraulic stability of natural circulation BWR remains unknown since increasing inlet restriction affect thermal-hydraulic stability due to reduction of the recirculation flow rate. Therefore experiments have been conducted to investigate the effect of inlet restriction on the thermal-hydraulic stability. A test facility used in this experiments was designed and constructed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation was described as a function of heat flux and inlet subcooling independent of inlet restriction. Stability maps in reference to the channel inlet subcooling, heat flux were presented for various inlet restriction which were carried out by an analysis based on the homogeneous flow various using this function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux. (author)

  15. Searching the beginning of BWR power instability events with the Hilbert Huang transform

    International Nuclear Information System (INIS)

    Blázquez, Juan; Montalvo, Cristina; García-Berrocal, Agustín; Balbás, Miguel

    2013-01-01

    Highlights: ► The report of the instability is enriched by including its beginning and its end. ► The Hilbert Huang transform (HHT) is used for indentifying both. ► The first Intrinsic Mode Function (IMF) detects both. ► The methodology is applied to neutron detector signals from two plants. ► The Decay Ratio of IMF 1 is calculated. - Abstract: When a BWR instability takes place, the Regulator usually demands a report which must include many aspects such as the initial time of the instability and also the measurements adopted by the operator at that time. This initial time normally is difficult to know from the available data. In this work, a methodology is proposed to determine accurately when the instability began based on the Hilbert–Huang transform. The Empirical Mode Decomposition is applied to neutron detector signals coming from two plants which have recorded them during real instability events. The first intrinsic mode function shows sharply the beginning and the end of the incident. Besides, through the instantaneous amplitude and frequency of the first mode a kind of Decay Ratio can be assigned allowing us to obtain a sharper description of the instability

  16. Siemens experience on linear and nonlinear analyses of out-of-phase BWR instabilities

    International Nuclear Information System (INIS)

    Kreuter, D.; Wehle, F.

    1995-01-01

    The Siemens design code STAIF has been applied extensively for linear analysis of BWR instabilities. The comparison between measurements and STAIF calculations for different plants under various conditions has shown good agreement for core-wide and regional instabilities. Based on the high quality of STAIF, the North German TUeV has decided to replace the licensing requirement of extensive stability measurements by predictive analyses with the code STAIF. Nonlinear stability analysis for beyond design boundary conditions with RAMONA has shown dryout during temporarily reversed flow at core inlet in case of core-wide oscillations. For large out-of-phase oscillations, dryout occurs already for small, still positive channel inlet flow. (orig.)

  17. Modal-based reduced-order model of BWR out-of phase instabilities

    International Nuclear Information System (INIS)

    Turso, J.A.; Edwards, R.M.; March-Leuba, J.

    1995-01-01

    For the past 40 yr, reduced-order modeling of boiling water reactor (BWR) dynamic behavior has been accomplished by several researchers. These models have been primarily concerned with providing insight into the so-called corewide neutron flux oscillation, where the power at each radial location in the core oscillates in unison. This is generally considered to be an illustration of the fundamental neutronic mode excited by the core thermal hydraulics. The time dependence of the fundamental mode is typically described by the point-kinetics equations, with one or more delayed-neutron groups. Thermal-hydraulic excitation of the first azimuthal harmonic mode, the so-called out-of-phase (OOP) instability, has been observed in operating BWRs. The temporal behavior of a low-order model of this phenomenon can be characterized using the modal point-kinetics formulation developed in this paper

  18. An analysis of instabilities of nuclear-coupled density-wave in BWR using modern frequency-domain control theory

    International Nuclear Information System (INIS)

    Zhao Yangping; Gao Huahun; Fu Longzhou

    1991-01-01

    A state-of-the-art multi-variable frequency-domain model has been developed for analysis of instabilities of nuclear-coupled density-wave in BWR core. The characteristic locus method is used for analysing the stability of BWR. A computer code-NUCTHIA has been derived. The model has been tested against the existing experimental data and compared with results of past single-variable analyses. By using the NUCTHIA code, the investigations of effects of main system parameters on BWW core stability have also been made. All the results are consistent with the experimental data

  19. The Physical Mechanism of Core-Wide and Local Instabilities at the Forsmark-1 BWR

    International Nuclear Information System (INIS)

    Analytis, G. Th.

    1998-10-01

    During the last 15 years, the problem of BWR instabilities has attracted the attention of a number of researchers. From the theoretical point of view, one would be interested in physically understanding the mechanisms responsible for the in- and out-of-phase core wide power oscillations observed at certain operating points of the power-flow map in different BWRs. From the practical point of view, one must try to avoid these 'incidents' since either locally, or globally, the power may substantially exceed the prescribed levels. In this work, we shall use RAMONA3-12 and analyse a rather unusual instability incident at Forsmark-1 in which in addition to the core-wide fundamental spatial mode oscillation, there were local large amplitude power oscillations at different radial positions in the core. We were able to reproduce these unusual experimental findings by assuming that there are large amplitude Density Wave Oscillations (DWOs) in different bundles, induced by the fact that these bundles were not seated properly into the lower fuel support plate. (author)

  20. Localisation of a channel instability in a BWR via neutron noise methods

    International Nuclear Information System (INIS)

    Karlsson, J.K.H.; Pazsit, I.

    1998-01-01

    A special type of instability occurred in the Swedish BWR Forsmark 1 in 1996. In contrast to the better known global or regional (out-of-phase) instabilities, the decay ratio appeared to be very high in one half of the core and quite low in the other half. A more detailed analysis showed that the most likely reason for the observed behaviour is a local perturbation of thermohydraulic character, e.g. a density wave oscillation (DWO), induced by the incorrect positioning of a fuel assembly (an 'unseated' assembly). In such a case it is of large importance to determine the position of the unseated assembly already during operation such that it can be easily found during reloading. The subject of this paper is to report on development and application of methods by which the position of such a local perturbation can be determined. Two different methods that support and complement each other were used. First a visualisation technique was elaborated which expedites a very good qualitative comprehension of the situation and which can be useful for the operators. It also gives an important basis for the application of the localisation algorithm. Second, a quantitative (algorithmic) localisation method, suited for this type of perturbation, was elaborated. This latter takes noise spectra from selected detectors as input and yields the perturbation position as output. The method was tested on simulated data, and then applied to the Forsmark measurements. The location of the disturbance, found by the algorithm, is in accordance with independent judgements for the case, and close to a position where an unseated assembly was found during refuelling. (author)

  1. Remark on the role of the driving force in BWR instability

    International Nuclear Information System (INIS)

    Dykin, V.; Pazsit, I.

    2009-01-01

    Simple models of BWR instability, used e.g. in understanding the role of the various oscillation modes in the overall stability of the plant, assume that each oscillation mode can be described by a second order system (a damped harmonic oscillator) driven by a white noise driving force. Change of the decay ratio (DR) of the observed signal is, as a rule, associated with the changing of the parameters of the damped oscillator, mainly its damping coefficient, and is interpreted in terms of the change of the stability of the system. However, conceptually, one cannot exclude cases when the change of the response of a driven damped oscillator is due to the change of the properties of the driving force. In this work we investigate the effect of a non-white driving force on the behaviour of the system. A question of interest is how changes of the spectrum of the driving force influence the observed autocorrelation function (ACF) of the resulting signal. Hence we calculate the response of a damped harmonic oscillator driven by a non-white driving force, corresponding to the reactivity effect of propagating density fluctuations in two-phase flow. It is shown how in some special cases such a driving force, when interpreting the neutron noise as if induced by a white noise driving source, can lead to an erroneous conclusion regarding the stability of the system. It is also concluded that in the practically interesting cases the effect of the coloured driving force, arising from propagating density fluctuations, is negligible.

  2. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  3. Application of Continuous and Structural ARMA modeling for noise analysis of a BWR coupled core and plant instability event

    International Nuclear Information System (INIS)

    Demeshko, M.; Dokhane, A.; Washio, T.; Ferroukhi, H.; Kawahara, Y.; Aguirre, C.

    2015-01-01

    Highlights: • We demonstrate the first application of a novel CSARMA method. • We analyze the instability occurred in a Swiss BWR plant during power ascension. • Benchmarked the results against STP analysis. • The CSARMA results are consistent with the background physics and the STP results. • The instability was caused by disturbances in the pressure control system. - Abstract: This paper presents a first application of a novel Continuous and Structural Autoregressive Moving Average (CSARMA) modeling approach to BWR noise analysis. The CSARMA approach derives a unique representation of the system dynamics by more robust and reliable canonical models as basis for signal analysis in general and for reactor diagnostics in particular. In this paper, a stability event that occurred in a Swiss BWR plant during power ascension phase is analyzed as well as the time periods that preceded and followed the event. Focusing only on qualitative trends at this stage, the obtained results clearly indicate a different dynamical state during the unstable event compared to the two other stable periods. Also, they could be interpreted as pointing out a disturbance in the pressure control system as primary cause for the event. To benchmark these findings, the frequency-domain based signal transmission-path (STP) method is also applied. And with the STP method, we obtained similar relationships as mentioned above. This consistency between both methods can be considered as being a confirmation that the event was caused by a pressure control system disturbance and not induced by the core. Also, it is worth noting that the STP analysis failed to catch the relations among the processes during the stable periods, that were clearly indicated by the CSARMA method, since the last uses more precise models as basis

  4. Development of a methodology of analysis of instabilities in BWR reactors; Desarrollo de una metodologia de analisis de inestabilidades en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

    This paper presents a methodology of analysis of the reactors instabilities of BWR type. This methodology covers of modal analysis of the point operation techniques of signal analysis and simulation of transients, through 3D Coupled RELAP5/PARCSv2.7 code.

  5. 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.

  6. A study on the feasibility of minor actinides in BWR

    International Nuclear Information System (INIS)

    Abdul Waris; Budiono

    2008-01-01

    Preliminary study on the feasibility of actinides minor (MA) recycling without mixing them with plutonium in boiling water reactor (BWR) has been carried out. The results show that increasing of fissile MA content in mixed oxide fuel (MOX) and/or reducing void fraction can enlarge the effective multiplication factor at the beginning of cycle, but the reactor still can not obtain its criticality condition. Furthermore, dropping the void fraction results in higher reactivity swing and therefore plummeting the safety factor of the reactor. (author)

  7. A study of heat capacity temperature limit of BWR

    International Nuclear Information System (INIS)

    Wang, Shih-Jen; Chen, Jyh-Jun; Chien, Chun-Sheng; Teng, Jyh-Tong

    2012-01-01

    Highlights: ► The purpose of this study is to verify the HCTL. ► MAAP4 was used as code to generate a realistic and convenient HCTL. ► The current HCTL curve causes confusing in reading data. ► The revised HCTL curves developed in this study. ► Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners’ group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

  8. A study of heat capacity temperature limit of BWR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shih-Jen, E-mail: sjenwang@iner.gov.tw [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Chen, Jyh-Jun [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China); Chien, Chun-Sheng [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer The purpose of this study is to verify the HCTL. Black-Right-Pointing-Pointer MAAP4 was used as code to generate a realistic and convenient HCTL. Black-Right-Pointing-Pointer The current HCTL curve causes confusing in reading data. Black-Right-Pointing-Pointer The revised HCTL curves developed in this study. Black-Right-Pointing-Pointer Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners' group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

  9. Studies of fragileness in steels of vessels of BWR reactors

    International Nuclear Information System (INIS)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2003-01-01

    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 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

  10. Study of transient turbine shot without bypass in a BWR

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    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)

  11. BWR Mark I pressure suppression study: bench mark experiments

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1977-01-01

    Computer simulations representative of the wetwell of Mark I BWR's have predicted pressures and related phenomena. However, calculational predictions for purposes of engineering decision will be possible only if the code can be verified, i.e., shown to compute in accord with measured values. Described in the report is a set of single downcomer spherical flask bench mark experiments designed to produce quantitative data to validate various air-water dynamic computations; the experiments were performed since relevant bench mark data were not available from outside sources. Secondary purposes of the study were to provide a test bed for the instrumentation and post-experiment data processing techniques to be used in the Laboratory's reactor safety research program and to provide additional masurements for the air-water scaling study

  12. Using the Hurst's exponent as a monitor and predictor of BWR reactor instabilities

    International Nuclear Information System (INIS)

    Gavilan Moreno, Carlos J.

    2010-01-01

    Since the decade of the 1950s, when the development of boiling water reactor technology began, unstable situations have existed, which involve a high amplitude self-oscillatory process in the reactor's thermal power. As the development progressed and the reactors increased power density, the possibility of instability under certain circumstances increased. Thus, in 1985, Caorso nuclear plant (Italy) reported the first event of this type, and in 1988, such an event was reported at La Salle as well. Since then, multiple instability events have been reported. The danger of these unstable power situations resides in the possibility of exceeding a thermal limit, as expressed in Appendix A of 10FR50. Thus, the need arises to monitor and correct these situations in the industry. The most common way to monitor and control these instability situations involves the use of Decay Ratio (DR) and Resonance Frequency (RF). The use of these parameters is polemical, because their use involves certain simplifications and operations prior to the calculation which question how well they represent the reality. The most important simplifications are those which lead to the interpretation of the power time series as the result of a second order system. With regard to the previous operations, the time series needs to be standardized and filtered. The result is loss of information during prediction, due to the operations, and the results, therefore, lack accuracy. In this paper, the system is considered without simplifications, that is to say that it is treated as dynamic and, as we shall see, chaotic, in the mathematical sense of the term. The series will be used in pure form without manipulations. The parameter used for monitoring and prediction of the core's behaviour will be the Hurst's exponent (H). The concept used for this proposal is that the response of a complex dynamic system depends not only on the last excitation, but on the prior history. The processes and systems are

  13. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1992-03-01

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  14. 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)

  15. Study on thermal performance and margins of BWR fuel elements

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

    This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

  16. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  17. 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

  18. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

    2008-01-01

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  19. A study on radiolytic nitrogen compounds in bwr primary systems

    International Nuclear Information System (INIS)

    Ibe, Eishi; Karasawa, Hidetoshi; Endo, Masao; Suzuki, Kazumichi; Etho, Yoshinori

    1988-01-01

    Behavior of nitrogen compounds in a γ radiation field was evaluated. Twenty-four species and 73 reactions were proposed for analysis of the system. It was pointed out that reactions dominating analysis reliability were primary reactions which lead to evolution of atomic nitrogen and reactions related to ammonium ion decomposition. Theoretical calculations for the BWR primary system revealed that: (i) in-leaked nitrogen from a condenser did not deteriorate the oxygen reduction efficiency due to hydrogen addition in reactor water; (ii) most 16 N atmos released in the main steam line were in the form of nitrogen mono-oxide under both hydrogen and normal water chemistry; (iii) 16 N atoms in nitric and nitrous acids under normal water chemistry were reduced by hydrogen atom to 16 NO and then released to the main steam line under hydrogen water chemistry; and (iv) 16 N in the main steam under normal water chemistry could be suppressed one order of magnitude by addition of non-radioactive nitrous acid into the reactor water. (author)

  20. Experimental study on reduced moderation BWR with Advanced Recycle System (BARS)

    International Nuclear Information System (INIS)

    Hiraiwa, K.; Yoshioka, K.; Yamamoto, Y.; Akiba, M.; Yamaoka, M.; Abe, N.; Mimatsu, J.

    2004-01-01

    Experimental study has been done for reduced-moderation spectrum boiling water reactor named BARS (BWR with Advanced Recycle System). The critical assembly experiment for triangular tight uranium lattice has been done in TOSHIBA critical assembly (NCA). Experimental method based on modified conversion ratio was adopted to evaluate the void reactivity effect. Void fraction was simulated by formed polystyrene in this experiment. The measured void coefficient for tight uranium lattice agreed with calculation. The thermal hydraulic test study has been done to study the coolability of BARS lattice. Visual test and high-pressure thermal hydraulic test have been done as the thermal hydraulic test. Visual test has indicated the flow behavior for BARS lattice is same as that of current BWR. The high-pressure thermal hydraulic test has indicated the applicability of modified Arai's correlation to the BARS lattice. (authors)

  1. Studies on core melt behaviour in a BWR pressure vessel lower head

    International Nuclear Information System (INIS)

    Lindholm, I.; Ikonen, K.; Hedberg, K.

    1999-01-01

    Core debris behaviour in the Nordic BWR lower head was investigated numerically using MELCOR and MAAP4 codes. Lower head failure due to penetration failure was studied with more detailed PASULA code taking thermal boundary conditions from MELCOR calculations. Creep rupture failure mode was examined with the two integral codes. Also, the possibility to prevent vessel failure by late reflooding was assessed in this study. (authors)

  2. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

    2001-01-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  3. Numerical study of jets secondary instabilities

    International Nuclear Information System (INIS)

    Brancher, Pierre

    1996-01-01

    The work presented in this dissertation is a contribution to the study of the transition to turbulence in open shear flows. Results from direct numerical simulations are interpreted within the framework of hydrodynamic stability theory. The first chapter is an introduction to the primary and secondary instabilities observed in jets and mixing layers. The numerical method used in the present study is detailed in the second chapter. The dynamics of homogeneous circular jets subjected to stream wise and azimuthal perturbations are investigated in the third chapter. A complete scenario describing the evolution of the jet is proposed with emphasis on the dynamics of vorticity within the flow. In the fourth chapter a parametric study reveals a three-dimensional secondary instability mainly controlled in the linear regime by the Strouhal number of the primary instability. In the nonlinear regime the dynamics of the azimuthal harmonies are described by means of model equations and are linked to the formation of stream wise vortices in the braid. The fifth chapter is dedicated to the convective or absolute nature of the secondary instabilities in plane shear layers. It is shown that there are flow configurations for which the two-dimensional secondary instability (pairing) is absolute even though the primary instability (Kelvin-Helmholtz) is convective. Some preliminary results concerning the three-dimensional secondary instabilities arc presented at the end of this chapter. The last chapter summarizes the main results and examines possible extensions of this work. (author) [fr

  4. Study on vertical seismic response model of BWR-type reactor building

    International Nuclear Information System (INIS)

    Konno, T.; Motohashi, S.; Izumi, M.; Iizuka, S.

    1993-01-01

    A study on advanced seismic design for LWR has been carried out by the Nuclear Power Engineering Corporation (NUPEC), under the sponsorship of the Ministry of International Trade and Industry (MITI) of Japan. As a part of the study, it has been investigated to construct an accurate analytical model of reactor buildings for a seismic response analysis, which can reasonably represent dynamic characteristics of the building. In Japan, vibration models of reactor buildings for horizontal ground motion have been studied and examined through many simulation analyses for forced vibration tests and earthquake observations of actual buildings. And now it is possible to establish a reliable horizontal vibration model on the basis of multi-lumped mass and spring model. However, vertical vibration models have not been so much studied as horizontal models, due to less observed data for vertical motions. In this paper, the vertical seismic response models of a BWR-type reactor building including soil-structure interaction effect are numerically studied, by comparing the dynamic characteristics of (1) three dimensional finite element model, (2) multi-stick lumped mass model with a flexible base-mat, (3) multi-stick lumped mass model with a rigid base-mat and (4) single-stick lumped mass model. In particular, the BWR-type reactor building has the long span truss roof which is considered to be one of the critical members to vertical excitation. The modelings of the roof trusses are also studied

  5. Analytic study of resistive instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Svensson, Magnus

    2003-05-01

    In a fusion plasma there is always a small amount of resistivity that may cause instabilities. Although their rather slow growth rates they can be of major importance for fusion plasma confinement. In this work a MAPLE-code was rewritten and simplified to make it possible to analytically solve the linearized MHD-equations with resistivity in an RFP-configuration. By using the MHD-equations and expanding the unknown perturbed quantities u{sub 1r}(r) and B{sub 1r}(r) as Taylor series and solving each coefficient we could get eigenvalues, dispersion relations and a relation between the growth rate and the resistivity. The new code was first used to solve two cases with no resistivity and simple unstable equilibria which after running gave the correct expected results. The difference from running the original code with these two cases was the greater speed of the calculations and the less memory needed. Then by using an ideal MHD-stable equilibrium in a plasma with no resistivity the code gave us solutions which unfortunately were not of the expected kind but the time of the calculations was still very fast. The resistivity was finally added to the code with the ideal MHD-stable equilibrium. The program also this time gave incorrect results. We could, however, see from a relation between the growth rate and the resistivity that the solution may be approximately correct in this domain. Although we did not get all the correct results we have to consider the fact that we got results, that were not possible before. Before this work was carried out we could not get any results at all in the resistive cue because of the very long memory demanding expressions. In future work and studies it is not only possible to get the desired eigenvalues {gamma} as function of {eta} but also possible to get expressions for eigenfunctions, dispersion relations and other significant relations with a number of variable parameters. We could also use the method for any geometry and possibly for

  6. Analytic study of resistive instabilities

    International Nuclear Information System (INIS)

    Svensson, Magnus

    2003-05-01

    In a fusion plasma there is always a small amount of resistivity that may cause instabilities. Although their rather slow growth rates they can be of major importance for fusion plasma confinement. In this work a MAPLE-code was rewritten and simplified to make it possible to analytically solve the linearized MHD-equations with resistivity in an RFP-configuration. By using the MHD-equations and expanding the unknown perturbed quantities u 1r (r) and B 1r (r) as Taylor series and solving each coefficient we could get eigenvalues, dispersion relations and a relation between the growth rate and the resistivity. The new code was first used to solve two cases with no resistivity and simple unstable equilibria which after running gave the correct expected results. The difference from running the original code with these two cases was the greater speed of the calculations and the less memory needed. Then by using an ideal MHD-stable equilibrium in a plasma with no resistivity the code gave us solutions which unfortunately were not of the expected kind but the time of the calculations was still very fast. The resistivity was finally added to the code with the ideal MHD-stable equilibrium. The program also this time gave incorrect results. We could, however, see from a relation between the growth rate and the resistivity that the solution may be approximately correct in this domain. Although we did not get all the correct results we have to consider the fact that we got results, that were not possible before. Before this work was carried out we could not get any results at all in the resistive cue because of the very long memory demanding expressions. In future work and studies it is not only possible to get the desired eigenvalues γ as function of η but also possible to get expressions for eigenfunctions, dispersion relations and other significant relations with a number of variable parameters. We could also use the method for any geometry and possibly for non

  7. Experimental study on low pressure flow instability

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin; Wu Shaorong; Bo Jinhai; Zhang Youjie

    1997-05-01

    The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The flow behavior for a wide range of inlet subcooling, in which the flow undergoes from single phase to two phase, is described in a natural circulation system at low pressure (p = 0.1, 0.24 MPa). Several kinds of flow instability, e.g. subcooled boiling instability, subcooled boiling induced flashing instability, pure flashing instability as well as flashing coupled density wave instability and high frequency flow oscillation, are investigated. The mechanism of flashing and flashing concerned flow instability, which has never been studied well in this field, is especially interpreted. The experimental results show that, firstly, for a low pressure natural circulation system the two phase flow is unstable in most of inlet subcooling conditions, the two phase stable flow can only be reached at very low inlet subcooling; secondly, at high inlet subcooling the flow instability is dominated by subcooled boiling in the heated section, and at middle inlet subcooling is dominated by void flashing in the adiabatic long riser; thirdly, in two phase stable flow region the condition for boiling out of the core, namely, single phase flow in the heated section, two phase flow in the riser due to vapor flashing, can be realized. The experimental results are very important for the design and accident analysis of the vessel and swimming pool type natural circulation nuclear heating reactor. (7 refs., 10 figs., 1 tab.)

  8. 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)

  9. 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)

  10. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    2013-01-01

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  11. Validation study of core analysis methods for full MOX BWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  12. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, N.; Su' ud, Z.; Riyana, E. S. [Nuclear Physics and Biophysics Research Division Department of Physics - Institut Teknologi Bandung (ITB) Jalan Ganeca 10 Bandung 40132 (Indonesia)

    2012-06-06

    A preliminary design study for the utilization of thorium added with {sup 231}Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of {sup 233}U to {sup 231}Pa in burn-up process. Optimizations of the content of {sup 231}Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 {approx} 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  13. MHD instability studies in ISX-B

    International Nuclear Information System (INIS)

    Pare, V.K.; Dunlap, J.L.; Navarro, A.P.; Burris, R.D.

    1979-01-01

    MHD instabilities in Ohmically and beam heated ISX-B plasmas have been studied using collimated x-ray and Mirnov loop diagnostics. The diagnostic systems will be described and the instability signals will be illustrated for a variety of discharges. The latter will include those observed in connection with low and high β operation, density clamping, pellet injection, and deliberate introduction of toroidal field ripple

  14. Study of transient rod extraction failure without RBM in a BWR

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    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. Studies on the fission products behavior during dissolution process of BWR spent fuel

    International Nuclear Information System (INIS)

    Sato, K.; Nakai, E.; Kobayashi, Y.

    1987-01-01

    In order to obtain basic data on fission products behavior in connection with the head end process of fuel reprocessing, especially to obtain better understanding on undissolved residues, small scale dissolution studies were performed by using BWR spent fuel rods which were irradiated as monitoring fuel rods under the monitoring program for LWR fuel assembly performance entitled PROVING TEST ON RELIABILITY OF FUEL ASSEMBLY . The Zircaloy-2 claddings and the fuel pellets were subjected individually to the following studies on 1) release of fission products during dissolution process, 2) characterization of undissolved residues, and 3) analysis of the claddings. This paper presents comprehensive descriptions of the fission products behavior during dissolution process, based on detailed and through PIE conducted by JNFS under the sponsorship of MITI (Ministry of International Trade and Industry)

  16. Feasibility studies of computed tomography in partial defect detection of spent BWR fuel

    International Nuclear Information System (INIS)

    Levai, F.; Tikkinen, J.; Tarvainen, M.; Arlt, R.

    1990-10-01

    Feasibility studies were made for tomographic reconstruction of a cross-sectional activity distribution of a spent nuclear fuel assembly. The purpose was to determine the number of fuel rods (pins) and localize the positisons where pins are missing. The activity distribution map showing the locations of fuel rods in the assembly was reconstructed. The theoretical part of this work consists of simulation of image reconstruction based on theoretically calculated data from a reference assembly model. Evaluation of different image reconstruction techniques was made. Measurements were made in real facility conditions. Gamma radiation from an irradiated 8 x 8 - 1 BWR fuel assembly was measured through a narrow custom made collimator from different angles and positions. The measured data set was used as projections for reconstructing the activity profile of the assembly in cross-sectional plane

  17. Linear study of the precessional fishbone instability

    Science.gov (United States)

    Idouakass, M.; Faganello, M.; Berk, H. L.; Garbet, X.; Benkadda, S.

    2016-10-01

    The precessional fishbone instability is an m = n = 1 internal kink mode destabilized by a population of trapped energetic particles. The linear phase of this instability is studied here, analytically and numerically, with a simplified model. This model uses the reduced magneto-hydrodynamics equations for the bulk plasma and the Vlasov equation for a population of energetic particles with a radially decreasing density. A threshold condition for the instability is found, as well as a linear growth rate and frequency. It is shown that the mode frequency is given by the precession frequency of the deeply trapped energetic particles at the position of strongest radial gradient. The growth rate is shown to scale with the energetic particle density and particle energy while it is decreased by continuum damping.

  18. Application of noise analysis for the study of core local instability at Forsmark 1

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1997-10-01

    Core local instability was recently experienced at Forsmark 1 BWR. The event has been studied by applying noise analysis to data collected in January 1997 for the stability test. The result indicated that there was a region in the left corner of the core which was subject to instability due to neutronic and thermal-hydraulic coupling. The result of the noise analysis suggested two types of disturbance source, one in the vicinity of the detector string LPRM10 having resonant oscillation at 0.5 Hz and another relatively wide band noise in the neighbourhood of LPRM18. Three hypotheses have been examined as the possible cause, operational factor, abnormal fuel assembly, and wide band low frequency disturbance. Although the real cause has not been made clear from the noise analysis, it is likely that the operational factor played an important role as the cause. Further investigations are expected to be performed in the future. In order to detect the local instability it is important to have a stability monitor with a capability of monitoring a sufficient number of LPRMs so as to cover the whole core. This is important since local instability is a type of anomaly which should not occur during reactor operation

  19. Plasma instabilities multifrequency study in equatorial electrojet

    International Nuclear Information System (INIS)

    Hanuise, C.

    1983-01-01

    In this thesis, multifrequential HF coherent radar results are presented, in the field plasma instabilities in equatorial electrojet. In a first part, characteristics of the irregularities observed either at the 3 meter wavelength by VHF radars, either at other wavelengths during pinpoint experiments, or in-situ by probe rockets are recalled. Theoretical studies progressed and are presented, at the same time with these experimental observations: instability linear theory, non linear theories, HF radar specificity, and problems associated to HF waves propagation and refraction in ionosphere. Original experimental results from Ethiopia are gathered in the second part. Plasma instability has been studied in different geophysical conditions and Doppler spectra characteristics are presented for each one of them. These characteristics are completely different according to the various cases; they are also different according to wether observations are made during the day in normal conditions (electric field pointed to the east at the equator) or in counter-electrojet conditions (electric field pointed to the west). The last part is concerned with theoretical interpretation of the previous results. A comprehensive view of the instability physical mechanisms, according to the geophysical conditions encountered, has been allowed by our results, VHF radar measurements at Jicamarca, or in situ probe measurements on the whole. Irregularities study has been limited to the E region [fr

  20. Preliminary study on characteristics of equilibrium thorium fuel cycle of BWR

    International Nuclear Information System (INIS)

    Waris, A.; Kurniadi, R.; Su'ud, Z.; Permana, S.

    2007-01-01

    One of the main objectives behind the transuranium recycling ideas is not merely to utilize natural resource that is uranium much more efficiently, but to reduce the environmental impact of the radio-toxicity of the nuclear spent fuel. Beside uranium resource, there is thorium which has three times abundance compared to that of uranium which can be utilized as nuclear fuel. On top of that thorium is believed to have less radio-toxicity of spent fuel since its produce smaller amount of higher actinides compared to that of uranium. However, the studies on the thorium utilization in nuclear reactor in particular in light water reactors (LWR) are not performed intensively yet. Therefore, the aim of the present study is to evaluate the characteristics of thorium fuel cycle in LWR, especially boiling water reactor (BWR). To conduct the comprehensive investigations we have employed the equilibrium burnup model (1-3). The equilibrium burnup model is an alternative powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor1). We have employed 1368 nuclides in the equilibrium burnup calculation where 129 of them are heavy metals (HMs). This burnup code then is coupled with SRAC cell calculation code by using PIJ module to compose an equilibrium-cell burnup code. For cell calculation, 26 HMs, 66 fission products (FPs) and one pseudo FP have been utilized. The JENDL 3.2 library has been used in this study. References: 1. A. Waris and H. Sekimoto, 'Characteristics of several equilibrium fuel cycles of PWR', J. Nucl. Sci. Technol., 38, p.517-526, 2001 2. A. Waris, H. Sekimoto, and G. Kastchiev, Influence of Moderator-to-Fuel Volume Ratio on Pu and MA Recycling in Equilibrium Fuel Cycles of

  1. A proposal for accident management optimization based on the study of accident sequence analysis for a BWR

    International Nuclear Information System (INIS)

    Sobajima, M.

    1998-01-01

    The paper describes a proposal for accident management optimization based on the study of accident sequence and source term analyses for a BWR. In Japan, accident management measures are to be implemented in all LWRs by the year 2000 in accordance with the recommendation of the regulatory organization and based on the PSAs carried out by the utilities. Source terms were evaluated by the Japan Atomic Energy Research Institute (JAERI) with the THALES code for all BWR sequences in which loss of decay heat removal resulted in the largest release. Identification of the priority and importance of accident management measures was carried out for the sequences with larger risk contributions. Considerations for optimizing emergency operation guides are believed to be essential for risk reduction. (author)

  2. A Non-Linear Digital Computer Model Requiring Short Computation Time for Studies Concerning the Hydrodynamics of the BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reisch, F; Vayssier, G

    1969-05-15

    This non-linear model serves as one of the blocks in a series of codes to study the transient behaviour of BWR or PWR type reactors. This program is intended to be the hydrodynamic part of the BWR core representation or the hydrodynamic part of the PWR heat exchanger secondary side representation. The equations have been prepared for the CSMP digital simulation language. By using the most suitable integration routine available, the ratio of simulation time to real time is about one on an IBM 360/75 digital computer. Use of the slightly different language DSL/40 on an IBM 7044 computer takes about four times longer. The code has been tested against the Eindhoven loop with satisfactory agreement.

  3. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    International Nuclear Information System (INIS)

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken

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

    International Nuclear Information System (INIS)

    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

  5. Review of two-phase instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Han Ok; Seo, Han Ok; Kang, Hyung Suk; Cho, Bong Hyun; Lee, Doo Jeong

    1997-06-01

    KAERI is carrying out a development of the design for a new type of integral reactors. The once-through helical steam generator is important design features. The study on designs and operating conditions which prevent flow instability should precede the introduction of one-through steam generator. Experiments are currently scheduled to understand two-phase instability, evaluate the effect of each design parameter on the critical point, and determine proper inlet throttling for the prevention of instability. This report covers general two-phase instability with review of existing studies on this topics. The general classification of two phase flow instability and the characteristics of each type of instability are first described. Special attention is paid to BWR core flow instability and once-through steam generator instability. The reactivity feedback and the effect of system parameters are treated mainly for BWR. With relation to once-through steam generators, the characteristics of convective heating and dryout point oscillation are first investigated and then the existing experimental studies are summarized. Finally chapter summarized the proposed correlations for instability boundary conditions. (author). 231 refs., 5 tabs., 47 figs

  6. Comparative study of the hydrogen generation during short term station blackout (STSBO) in a BWR

    International Nuclear Information System (INIS)

    Polo-Labarrios, M.A.; Espinosa-Paredes, G.

    2015-01-01

    Highlights: • Comparative study of generation in a simulated STSBO severe accident. • MELCOR and SCDAP/RELAP5 codes were used to understanding the main phenomena. • Both codes present similar thermal-hydraulic behavior for pressure and boil off. • SCDAP/RELAP5 predicts 15.8% lower hydrogen production than MELCOR. - Abstract: The aim of this work is the comparative study of hydrogen generation and the associated parameters in a simulated severe accident of a short-term station blackout (STSBO) in a typical BWR-5 with Mark-II containment. MELCOR (v.1.8.6) and SCDAP/RELAP5 (Mod.3.4) codes were used to understand the main phenomena in the STSBO event through the results comparison obtained from simulations with these codes. Due that the simulation scope of SCDAP/RELAP5 is limited to failure of the vessel pressure boundary, the comparison was focused on in-vessel severe accident phenomena; with a special interest in the vessel pressure, boil of cooling, core temperature, and hydrogen generation. The results show that at the beginning of the scenario, both codes present similar thermal-hydraulic behavior for pressure and boil off of cooling, but during the relocation, the pressure and boil off, present differences in timing and order of magnitude. Both codes predict in similar time the beginning of melting material drop to the lower head. As far as the hydrogen production rate, SCDAP/RELAP5 predicts 15.8% lower production than MELCOR

  7. 3D analysis methods - Study and seminar[BWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Daaviittila, A [Valtion Teknillinen Tutkimuskeskus (Finland)

    2003-10-01

    The first part of the report results from a study that was performed as a Nordic co-operation activity with active participation from Studsvik Scandpower and Westinghouse Atom in Sweden, and VTT in Finland. The purpose of the study was to identify and investigate the effects rising from using the 3D transient com-puter codes in BWR safety analysis, and their influence on the transient analysis methodology. One of the main questions involves the critical power ratio (CPR) calculation methodology. The present way, where the CPR calculation is per-formed with a separate hot channel calculation, can be artificially conservative. In the investigated cases, no dramatic minimum CPR effect coming from the 3D calculation is apparent. Some cases show some decrease in the transient change of minimum CPR with the 3D calculation, which confirms the general thinking that the 1D calculation is conservative. On the other hand, the observed effect on neutron flux behaviour is quite large. In a slower transient the 3D effect might be stronger. The second part of the report is a summary of a related seminar that was held on the 3D analysis methods. The seminar was sponsored by the Reactor Safety part (NKS-R) of the Nordic Nuclear Safety Research Programme (NKS). (au)

  8. Beam instability studies at the APS

    International Nuclear Information System (INIS)

    Teng, L.C.

    1994-01-01

    The Argonne Advanced Photon Source, APS (Fig. 1), is a 7-GeV positron storage ring with a circumference of 1104 m. It has a ''third generation, DBA or Chasman-Green'' lattice composed of 40 sectors each having a ∼6 m long zero-dispersion straight-section for accommodating insertion devices. Neighboring straight-sections are connected by a 360 degrees/40 = 9 degrees double-bend-achromatic bending section designed to produce the smallest emittance attainable with reasonable component parameter values and dynamic apertures. Thus, it is a very strongly focusing lattice with v x = 35.22 and v y = 14.30. The beam chamber of the storage ring including all rf, vacuum and photon beam components is designed to ensure that a beam current > 100 mA can be stably stored. We expect that the maximum stable beam current could be as high as 300 mA. This paper will give some details of the studies and computations to ensure the stability of such a beam. The discussions will be organized in the following three parts: Coupled-bunch instability caused by the higher-order modes (HOMs) of the rf cavities; Single-bunch instability due to the resistive wall impedance; and Single-bunch instability due to broadband impedances arising from beam chamber irregularities

  9. Recent studies on the welding of austenitic stainless steel piping for BWR service

    International Nuclear Information System (INIS)

    Childs, W.J.

    1986-01-01

    The incidence of intergranular stress corrosion cracking (IGSCC) in stainless steel piping in BWR power plants has led to the development of various countermeasures. Replacement of the susceptible Type 304 stainless steel with Type 316 nuclear grade stainless steel has been done by a number of plants. In order to minimize radiation exposure to welding personnel, automatic GTA welding has been used wherever possible when we make the field welds. Studies have shown that the residual stresses in the welded butt joints are affected by the welding process, weld joint design and welding procedures. A new weld joint design has been developed which minimizes the volume of deposited metal while providing adequate access for welding. It also minimizes axial and radial shrinkage and the resulting residual stresses. Other countermeasures, which have been used, include stress modifications such as induction heating stress improvement (IHSI) and last pass heat sink welding (LPHSW). It has been shown that these remedies must be process adjusted to account for the welding process employed. In some cases where UT cracking indication have been detected or where through wall cracking has occurred, weld surfacing has been used to extend life. A further approach to preventing IGSCC in the weld HAZ has been through improvement of the water chemistry by injecting hydrogen to reduce the oxygen level and by keeping the impurity level low

  10. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    International Nuclear Information System (INIS)

    Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu

    1999-01-01

    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)

  11. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

  12. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study

    International Nuclear Information System (INIS)

    Salinas H, J.G.; Espinosa P, G.; Gonzalez M, V.M.

    2000-01-01

    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. 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.

  14. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Cruz G, M.; Amador C, C.

    2013-10-01

    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)

  15. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  16. Feasibility study on development of plate-type heat exchanger for BWR plants

    International Nuclear Information System (INIS)

    Ohyama, Nobuhiro; Suda, Kenichi; Ogata, Hiroshi; Matsuda, Shinichi; Nagasaka, Kazuhiro; Fujii, Toshi; Nozawa, Toshiya; Ishihama, Kiyoshi; Higuchi, Tomokazu

    2004-01-01

    In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

  17. Numerical MHD study for plasmoid instability in uniform resistivity

    Science.gov (United States)

    Shimizu, Tohru; Kondoh, Koji; Zenitani, Seiji

    2017-11-01

    The plasmoid instability (PI) caused in uniform resistivity is numerically studied with a MHD numerical code of HLLD scheme. It is shown that the PI observed in numerical studies may often include numerical (non-physical) tearing instability caused by the numerical dissipations. By increasing the numerical resolutions, the numerical tearing instability gradually disappears and the physical tearing instability remains. Hence, the convergence of the numerical results is observed. Note that the reconnection rate observed in the numerical tearing instability can be higher than that of the physical tearing instability. On the other hand, regardless of the numerical and physical tearing instabilities, the tearing instability can be classified into symmetric and asymmetric tearing instability. The symmetric tearing instability tends to occur when the thinning of current sheet is stopped by the physical or numerical dissipations, often resulting in the drastic changes in plasmoid chain's structure and its activity. In this paper, by eliminating the numerical tearing instability, we could not specify the critical Lundquist number Sc beyond which PI is fully developed. It suggests that Sc does not exist, at least around S = 105.

  18. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

  19. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  20. Application of autoregressive methods and Lyapunov coefficients for instability studies of nuclear reactors

    International Nuclear Information System (INIS)

    Aruquipa Coloma, Wilmer

    2017-01-01

    Nuclear reactors are susceptible to instability, causing oscillations in reactor power in specific working regions characterized by determined values of power and coolant mass flow. During reactor startup, there is a greater probability that these regions of instability will be present; another reason may be due to transient processes in some reactor parameters. The analysis of the temporal evolution of the power reveals a stable or unstable process after the disturbance in a light water reactor of type BWR (Boiling Water Reactor). In this work, the instability problem was approached in two ways. The first form is based on the ARMA (Autoregressive Moving Average models) model. This model was used to calculate the Decay Ratio (DR) and natural frequency (NF) of the oscillations, parameters that indicate if the one power signal is stable or not. In this sense, the DRARMA code was developed. In the second form, the problems of instability were analyzed using the classical concepts of non-linear systems, such as Lyapunov exponents, phase space and attractors. The Lyapunov exponents quantify the exponential divergence of the trajectories initially close to the phase space and estimate the amount of chaos in a system; the phase space and the attractors describe the dynamic behavior of the system. The main aim of the instability phenomena studies in nuclear reactors is to try to identify points or regions of operation that can lead to power oscillations conditions. The two approaches were applied to two sets of signals. The first set comes from signals of instability events of the commercial Forsmark reactors 1 and 2 and were used to validate the DRARMA code. The second set was obtained from the simulation of transient events of the Peach Bottom reactor; for the simulation, the PARCS and RELAP5 codes were used for the neutronic/thermal hydraulic coupling calculation. For all analyzes made in this work, the Matlab software was used due to its ease of programming and

  1. Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

    2002-01-01

    Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

  2. Symbol synchronizer assembly instability study, part 2

    Science.gov (United States)

    Bunce, R. C.

    1975-01-01

    Data processing to develop a third-order phase model and to translate all such processed data to the frequency realm for further analysis was described. The frequency study yields a long frequency modulation (FM) drift sinusoid (1600-sec period), an impressed secondary drift wave with a period of about 116 sec, and a set of even harmonics of twice the ramp period-the latter arising from, and used to modify, the phase detector model. The result is applied secondarily to estimate the strong-signal SSA phase detector response "out-of-lock." Finally, the main drift components are verified against all available data, and the result is used to estimate minimum lock conditions and the SSA drift effect under normal operating modes. The instability problem appears marginally resolvable if the acquisition technique is modified.

  3. Experimental study on breakup and fragmentation behavior of molten material jet in complicated structure of BWR lower plenum

    International Nuclear Information System (INIS)

    Saito, Ryusuke; Abe, Yutaka; Yoshida, Hiroyuki

    2014-01-01

    To estimate the state of reactor pressure vessel of Fukushima Daiichi nuclear power plant, it is important to clarify the breakup and fragmentation of molten material jet in the lower plenum of boiling water reactor (BWR) by a numerical simulation. To clarify the effects of complicated structures on the jet behavior experimentally and validate the simulation code, we conduct the visualized experiments simulating the severe accident in the BWR lower plenum. In this study, jet breakup, fragmentation and surrounding velocity profiles of the jet were observed by the backlight method and the particle image velocimetry (PIV) method. From experimental results using the backlight method, it was clarified that jet tip velocity depends on the conditions whether complicated structures exist or not and also clarified that the structures prevent the core of the jet from expanding. From measurements by the PIV method, the surrounding velocity profiles of the jet in the complicated structures were relatively larger than the condition without structure. Finally, fragment diameters measured in the present study well agree with the theory suggested by Kataoka and Ishii by changing the coefficient term. Thus, it was suggested that the fragmentation mechanism was mainly controlled by shearing stress. (author)

  4. Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

    Energy Technology Data Exchange (ETDEWEB)

    Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

    1976-07-01

    Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

  5. Experimental and analytical studies for a BWR nuclear reactor building. Evaluation of soil-structure interaction behaviour

    International Nuclear Information System (INIS)

    Mizuno, N.; Tsushima, Y.

    1975-01-01

    This paper evaluates the spatial characteristics of dynamic properties, especially soil-structure interaction behaviour, of the BWR nuclear building by experimental and analytical studies. It is well known that the damping effects in soil-structure interaction are remarkable on the building with short periods by the dissipation of vibrational energy to the soil. The authors have previously reported an analytical method for estimating the damping effects the properties of which are characterized as follows: 1) The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. 2) H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. In this paper, an approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. An approximate explanation is presented in regard to the experimental results of the No. 1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. (Auth.)

  6. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  7. Experimental study of the longitudinal instability for beam transport

    International Nuclear Information System (INIS)

    Reiser, M.; Wang, J.G.; Guo, W.M.; Wang, D.X.

    1990-01-01

    Theoretical model for beam longitudinal instability in a transport pipe with general wall impedance is considered. The result shows that a capacitive wall tends to stabilize the beam. The experimental study of the instability for a pure resistive-wall is presented, including the design parameters, setup and components for the experiment. 6 refs., 3 figs

  8. Study of genomic instability induced by low dose ionizing radiation

    International Nuclear Information System (INIS)

    Seoane, A.; Crudeli, C.; Dulout, F.

    2006-01-01

    The crews of commercial flights and services staff of radiology and radiotherapy from hospitals are exposed to low doses of ionizing radiation. Genomic instability includes those adverse effects observed in cells, several generations after the exposure occurred. The purpose of this study was to analyze the occurrence of genomic instability by very low doses of ionizing radiation [es

  9. Studies of elastic-plastic instabilities

    DEFF Research Database (Denmark)

    Tvergaard, Viggo

    1999-01-01

    Analyses of plastic instabilities are reviewed, with focus on results in structural mechanics as well as continuum mechanics. First the basic theories for bifurcation and post-bifurcation behavior are briefly presented. Then, localization of plastic flow is discussed, including shear band formati...

  10. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  11. Cobalt deposition studies in the primary circuit under BWR conditions (Phase 1 and 2)

    International Nuclear Information System (INIS)

    Bennett, Peter

    1996-04-01

    This report presents the results from the first 2 phases of an experiment performed to investigate the effects of water chemistry on cobalt transport and deposition in the primary circuit under BWR conditions. Two high pressure water loops have been used to compare the incorporation of cobalt into the oxide films on coupons of various LWR primary circuit constructional materials, with several pretreatments, under Hydrogen Water Chemistry (HWC) and Normal Water Chemistry (NWC) conditions. Cobalt-60 deposition rates onto samples that had been pre-oxidised in air were lower than on samples that had been exposed previously in a water loop or had untreated surfaces. In NWC, oxide layers were thicker, normalised Co-60 deposition rates were higher and Co-60 activities per unit volume of oxide were greater. Some evidence has been produced to support the conclusions of other workers that a chromium-rich outer oxide layer is responsible for enhanced cobalt incorporation. (author)

  12. Applicability study on a ceramic filter with hot-test conducted in a BWR plant

    International Nuclear Information System (INIS)

    Yamada, K.; Shirai, T.; Wada, M.; Nakamizo, H.

    1991-01-01

    Radioactive crud removal and filtration performance recovery by backwashing were examined with a BWR plant pool water using a ceramic filter element, 0.1 micron in nominal pore size and 0.2m 2 in filtration area. Totally 1114 hours filter operation were accumulated. Ten backwashings were accomplished during the test period. The following results were obtained. (1) Radioactive crud concentration in the filter effluent remained below 10 5 Bq/m 3 . (2) Both pressure loss through the filter and dose rate at the filter vessel surface were recovered to the initial level by each backwashing. The surface dose rate after backwashing was approximately 0.01mSv/h. According to these test results, it is confirmed that the ceramic filter is appropriate for the treatment of highly crud concentrated radioactive liquid, which is generated in nuclear facilities, such as spent fuel reprocessing plants. (author)

  13. OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

    International Nuclear Information System (INIS)

    Ivanov, Kostadin; Olson, Andy; Sartori, Enrico

    2004-01-01

    An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark

  14. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Porsmyr, Jan; Bodal, Terje; Beere, William H.

    2004-01-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

  15. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

    2004-07-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

  16. A study on the criticality search of transuranium recycling BWR core by adjusting supplied fuel composition in equilibrium state

    International Nuclear Information System (INIS)

    Seino, Takeshi; Sekimoto, Hiroshi

    1998-01-01

    There have been some difficulties in carrying out an extensive evaluation of the equilibrium state of Light Water Reactor (LWR) recycling operations keeping their fixed criticality condition using conventional design codes because of the complexity of their calculation model for practical fuel and core design and because of a large amount of calculation time. This study presents an efficient approach to secure the criticality in an equilibrium cycle by adjusting a supplied fuel composition. The criticality search is performed by the use of fuel importance obtained from the equation adjoint to a continuously fuel supplied core burnup equation. Using this method, some numerical analyses were carried out in order to evaluate the mixed oxide (MOX) fuel composition of equilibrium Boiling Water Reactor (BWR) cores satisfying the criticality requirement. The results showed the comprehensive and quantitative characteristics on the equilibrium cores confining transuraniums for different MOX fuel loading fractions and irradiating conditions

  17. A study of the effect of space-dependent neutronics on stochastically-induced bifurcations in BWR dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.T. [Paul Scherrer Institute (PSI), Villigen (Switzerland)

    1995-09-01

    A non-linear one-group space-dependent neutronic model for a finite one-dimensional core is coupled with a simple BWR feed-back model. In agreement with results obtained by the authors who originally developed the point-kinetics version of this model, we shall show numerically that stochastic reactivity excitations may result in limit-cycles and eventually in a chaotic behaviour, depending on the magnitude of the feed-back coefficient K. In the framework of this simple space-dependent model, the effect of the non-linearities on the different spatial harmonics is studied and the importance of the space-dependent effects is exemplified and assessed in terms of the importance of the higher harmonics. It is shown that under certain conditions, when the limit-cycle-type develop, the neutron spectra may exhibit strong space-dependent effects.

  18. A study on the criticality search of transuranium recycling BWR core by adjusting supplied fuel composition in equilibrium state

    International Nuclear Information System (INIS)

    Seino, Takeshi; Sekimoto, Hiroshi

    1997-01-01

    There have been some difficulties in carrying out an extensive evaluation of the equilibrium state of Light Water Reactor (LWR) recycling operations keeping their fixed criticality condition using conventional design codes, because of the complexity of their calculational model for practical fuel and core design and because of a large amount of calculation time. This study presents an efficient approach to secure the criticality in an equilibrium cycle by adjusting a supplied fuel composition. The criticality search is performed by the use of fuel importance obtained from the equilibrium adjoint to a continuously fuel supplied core burnup equation. Using this method, some numerical analyses were carried out in order to evaluate the mixed oxide (MOX) fuel composition of equilibrium Boiling Water Reactor (BWR) cores satisfying the criticality requirement. The results showed the comprehensive and quantitative characteristics on the equilibrium cores confining transuranium for different MOX fuel loading fractions and irradiating conditions. (author)

  19. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  20. Valuation of power oscillations in a BWR after control rod banks withdrawal events

    International Nuclear Information System (INIS)

    Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

    2009-01-01

    The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

  1. Studies of intense-laser plasma instabilities

    Czech Academy of Sciences Publication Activity Database

    Láska, Leoš; Krása, Josef; Badziak, J.; Jungwirth, Karel; Krouský, Eduard; Margarone, Daniele; Parys, P.

    2013-01-01

    Roč. 272, May (2013), 94-98 ISSN 0169-4332 R&D Projects: GA MŠk(CZ) 7E09092; GA MŠk(CZ) LC528; GA AV ČR IAA100100715 Institutional research plan: CEZ:AV0Z10100523 Keywords : laser plasma instabilities * self-generated magnetic field * longitudinal structure of the expanding plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.538, year: 2013

  2. Experimental and analytical studies for a BWR nuclear reactor building evaluation of soil-structure interaction behavior

    International Nuclear Information System (INIS)

    Mizuno, N.; Tsushima, Y.

    1975-01-01

    The purpose of this paper is to evaluate the spatial characteristics of dynamic properties, especially soil-structure interaction behavior, or the BWR nuclear reactor building by experimental and analytical studies. An analytical method (SMIRT-1 Paper K 2/4) for estimating the damping effects is reported. The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. An approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. Next, an approximate explanation is presented in regard to the experimental results of the No.1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. The regression analyses of the experimental resonance curves by one degree system show that the critical damping ratio is larger than the 0.10 used in the design for the fundamental natural period. It is attempted to simulate the experimental results by the above-mentioned method. The simulated model is a fourty-eight degrees of freedom spring mass system because of the eight masses for the eight floors including the base foundation and the six degrees of freedom for a mass

  3. Numerical studies of edge localized instabilities in tokamaks

    International Nuclear Information System (INIS)

    Wilson, H.R.; Snyder, P.B.; Huysmans, G.T.A.; Miller, R.L.

    2002-01-01

    A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code

  4. Utilization of a technical review group during a BWR owners group technical specification improvement study

    International Nuclear Information System (INIS)

    Hansell, H.F.; Moyer, D.P.

    1986-01-01

    A BWR Owners' Group Technical Specification Improvement (TSI) Committee was formed in late 1983. A primary goal of this Committee was to encourage the development of a probabilistic methodology for technical specification improvements which could be readily applied by utilities. The TSI Committee elected to hire a Contractor to develop and demonstrate a method. After the Contractor was selected and has started work, the committee decided to establish a Technical Review Group (TRG) to efficiently and effectively review the Contractor's analyses. The TRG met frequently with the Contractor as the analyses were being performed. These meetings were held at the Contractor's facility in order to allow direct contact between reviewers and individuals performing the work. The TRG was also involved with all major interactions with the NRC. The significance and merit of using a peer review group in this manner is the theme of this paper. In order to present a discussion of the significance and merit of the TRG, the activities are described. The summary of the analytical approach is provided to more full understand the TRG activities

  5. Experimental study of advanced continuous acoustic emission monitoring of BWR components. Final report

    International Nuclear Information System (INIS)

    McElroy, J.W.; Hartman, W.F.

    1980-09-01

    The program consisted of installing, maintaining, and monitoring AE sensors located on primary piping, nozzles, and valves in the BWR system. Analysis of the AE data was correlated to the results of supplementary nondestructive testing techniques used during the in-service inspection, performed at refueling outages. Purpose of the program was to develop the on-line surveillance acoustic emission technique in order to identify areas of possible structural degradation. Result of reducing inspection time was to reduce accumulated radiation exposure to inspecting personnel and to reduce the amount of critical plant outage time by identifying the critical inspection areas during operation. The program demonstrated the capability of acoustic emission instrumentation to endure the nuclear reactor environment. The acoustic emission sensors withstood 12 months of reactor operation at temperatures of 400 0 F and greater in high radiation fields. The preamplifiers, also mounted in the reactor environment, operated for the 12-month period in 100% humidity, 250 0 F conditions. The remaining cable and AE instrumentation were operated in controlled environments

  6. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  7. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  8. Basic study on BWR plant behavior under the condition of severe accident

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Jyohko, Shingo; Dohgo, Hirofumi

    2015-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the severe accident that LOCA occurs but ECCS fails the water irrigation into the reactor core. The simulation experiments were carried out for the cases that LOCA has occurred in the main steam piping or in the recirculation piping, respectively. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for LOCA area of up to 100 cm"2 for both type LOCA. In the case of main steam system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the recirculation system LOCA of 2000 cm"2 area, a drop of the core water level was extremely in comparison with main steam system LOCA, and the fuel assemblies were completely exposed during up to 30 min, to the irrigation from approximately 100 sec, after LOCA occurrence. Therefore, the fuel temperature during the irrigation had reached approximately 1900degC. Thus, the fuel cladding were damaged approximately less than 10%, and H_2 concentration in the containment vessel was approximately 9% which did not exceed H_2 detonation limit of 13% but exceeded H_2 explosion limit of 4%. However, the containment vessel internal pressure was settled around design pressure value of containment vessel. As the results, some core damage could not be avoided, but soundness of the containment vessel, which should take the role of 'confine', was found to be secured. (author)

  9. Basic study on BWR plant behavior under the condition of severe accident (2)

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Ueda, Masataka; Sasaki, Hajime

    2016-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the two types of severe accidents that LOCA occurs but ECCS fails the water irrigation into the reactor core and SBO occurs and at the same time the reclosed failure of SRV occurs. The simulation experiments were carried out for the cases that LOCA has occurred in the main feed-water piping. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for small and middle LOCA area. In the case of main feed-water system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the accident of SBO with reclosed failure of SRV, it has been shown that the accidents continue to progress rapidly as compared with the case of normally operating of SRV. Because SRV has the function that keep the inside pressure of reactor core by repeating opened and closed in response of the inside pressure and prevent the decrease of water level inside reactor core. However, if the irrigation into the reactor core was carried out 30 min after SBO occurrence, the core had little damage and also the H_2 concentration in the containment vessel did not exceed limits of H_2 explosion. Further, as for the accident of reclosed failure of SRV, it has been shown that there are very good correspondence with the simulation results of main steam piping LOCA of area 180 cm"2 corresponding to the inlet cross-sectional area SRV installed on the piping

  10. Threshold Studies of the Microwave Instability in Electron Storage Rings

    International Nuclear Information System (INIS)

    Bane, Karl

    2010-01-01

    We use a Vlasov-Fokker-Planck program and a linearized Vlasov solver to study the microwave instability threshold of impedance models: (1) a Q = 1 resonator and (2) shielded coherent synchrotron radiation (CSR), and find the results of the two programs agree well. For shielded CSR we show that only two dimensionless parameters, the shielding parameter Π and the strength parameter S csr , are needed to describe the system. We further show that there is a strong instability associated with CSR, and that the threshold, to good approximation, is given by (S csr )th = 0.5 + 0.12Π. In particular, this means that shielding has little effect in stabilizing the beam for Π ∼ -3/2 . We, in addition, find another instability in the vicinity of Π = 0.7 with a lower threshold, (S csr ) th ∼ 0.2. We find that the threshold to this instability depends strongly on damping time, (S csr ) th ∼ τ p -1/2 , and that the tune spread at threshold is small - both hallmarks of a weak instability.

  11. Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

    2003-01-01

    Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

  12. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  13. Studies of thermal-hydrodynamic flow instability, (3)

    International Nuclear Information System (INIS)

    Suzuoki, Akira

    1978-01-01

    In the flow system in which large density change occurs midway, sometimes steady flow cannot be maintained according to the conditions, and pulsating flow or the scamper of flow occurs. This phenomenon is called flow instability, and is noticed as one of the causes to obstruct the normal operation in boilers, BWRs and the steam generators for FBRs with parallel evaporating tube system. In the pulsating instability, there are density wave oscillation and pressure wave oscillation. The author has studied the density wave oscillation occurring in the steam generators for FBRs and in this paper, the role played by two-phase flow regarding the occurrence of flow instability, and the effect of the existence of interphase slip on the role played by two-phase flow are reported. The theoretical analysis and the results of the analysis taking a steam generator heated with sodium as the example are described. Regarding flow stability, two-phase flow part generates the variation of weight velocity with different phase in steam single phase part, accepting enthalpy variation in water single phase part. In this action, the effect of interphase slip was observed, and the variation of reverse phase is apt to occur in slip flow as compared with homogeneous flow. Accordingly, flow instability is apt to occur in slip flow. (Kako, I.)

  14. Study of the possibility of localising a channel instability in Forsmark-1

    International Nuclear Information System (INIS)

    Karlsson, J.K-H.; Pazsit, I.

    1998-01-01

    A special type of instability occurred in the Swedish BWR Forsmark 1 in 1996. In contrast to the better known global or regional instabilities, the decay ratio appeared to be very high in one half of the core and quite low in the other half. A more detailed analysis showed that the most likely reason for the observed behaviour is a local perturbation of thermohydraulic character induced by the incorrect positioning of a fuel assembly. In such a case it is of importance to determine the position of the unseated assembly already during operation so that it can be easily found during reloading. The subject of this paper is to report on development and application of a method by which the position of such a local perturbation can be determined. The method can be separated into two parts that support and complement each other. First a visualisation technique was elaborated which displays the space-time behaviour of the neutron flux oscillations in the core. This visualisation expedites a very good qualitative comprehension of the situation and can be useful for the operators. It also gives a basis for the application of the localisation algorithm. Second, a quantitative localisation method, based on reactor physical models of the perturbation and of the transfer function between the perturbation and the flux oscillations, was elaborated. This latter takes noise spectra from selected detectors as input and yields the perturbation position as output. The strength of the method lies in its potentially high spatial resolution, which is smaller than the typical distance between two adjacent LPRM detectors. The method was tested on simulated data, and then applied to the Forsmark measurements. The location of the disturbance, found by the algorithm, is in accordance with independent judgements for the case, and close to a position where an unseated assembly was found during refuelling. The purpose of this study was to develop and test the localisation method. To apply the

  15. Theoretical study on the first kind of density wave instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Zuying, Gao; Jincai, Li; Baocheng, Xu; Zuoyi, Zhang; Cheng, Gao [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The present paper summarizes the theoretical studies carried out by INET (Institute of Nuclear Energy Technology) of Tsinghua University on the first kind of density wave instabilities (DWIs) of natural circulation systems. The analysis methods of DWI and mathematical models of drift flux are presented. Based on the general excess entropy production criterion of non-equilibrium thermodynamics, an energy principle of DWI is established. (author). 10 refs, 16 figs.

  16. Study of current instabilities in high resistivity gallium arsenide

    International Nuclear Information System (INIS)

    Barraud, A.

    1968-01-01

    We have shown the existence and made a study of the current oscillations produced in high-resistivity gallium arsenide by a strong electric field. The oscillations are associated with the slow travelling of a region of high electrical field across the whole sample. An experimental study of the properties of these instabilities has made it possible for us to distinguish this phenomenon from the Gunn effect, from acoustic-electric effects and from contact effects. In order to account for this type of instability, a differential trapping mechanism involving repulsive impurities is proposed; this mechanism can reduce the concentration of charge carriers in the conduction band at strong electrical fields and can lead to the production of a high-field domain. By developing this model qualitatively we have been able to account for all the properties of high-resistance gallium arsenide crystals subjected to a strong electrical field: increase of the Hall constant, existence of a voltage threshold for these oscillations, production of domains of high field, low rate of propagation of these domains, and finally the possibility of inverting the direction of the propagation of the domain without destroying the latter. A quantitative development of the model makes it possible to calculate the various characteristic parameters of these instabilities. Comparison with experiment shows that there is a good agreement, the small deviations coming especially from the lack of knowledge concerning transport properties in gallium arsenide subjected to high fields. From a study of this model, it appears that the instability phenomenon can occur over a wide range of repulsive centre concentrations, and also for a large range of resistivities. This is the reason why it appears systematically in gallium arsenide of medium and high resistivity. (authors) [fr

  17. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  18. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

    International Nuclear Information System (INIS)

    Ojeda S, J.; Morales S, J.; Chavez M, C.

    2009-10-01

    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)

  19. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  20. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  1. Experimental studies of the effect of functional spacers to annular flow in subchannels of a BWR fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Damsohn, M., E-mail: damsohn@lke.mavt.ethz.c [ETH Zurich, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zuerich (Switzerland); Prasser, H.-M. [ETH Zurich, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zuerich (Switzerland)

    2010-10-15

    For the prediction of dryout in fuel elements of boiling water reactors, the dynamic behavior of the water film covering the fuel rod has to be understood. This paper provides high resolved experimental data of the liquid film and gives insight into the dynamic film behavior. The experiments of this work were conducted in a vertical channel representing a pair of neighboring subchannels of a BWR fuel rod bundle. Air and water at ambient pressure and temperature are used as model fluids, creating an annular flow in the test section. The influence of different functional spacer shapes on the liquid film has been studied. The heart of the instrumentation is a liquid film sensor (LFS), which measures the film thickness distribution around a half cylinder with a matrix of 64 x 16 measuring points with a time resolution of 10,000 frames per second and a spatial resolution of 2 mm x 2 mm. The high resolution allows for a visualization of the dynamic liquid film as a movie animation. Principals of the dynamic behavior of the liquid film are observed. The time-averaged film thickness distributions show that the spacers structure the liquid film significantly. The gaseous phase is accelerated due to the cross-section blockage caused by the spacer. This leads to a local thinning of the liquid film downstream of the spacer. Two statistical evaluation methods are presented to determine different dynamic wave properties: The wave velocity as function of the wave height, the traveling path of the waves and the location of wave separation and merge events. The first evaluation method shows that big waves move generally faster than small waves. The analysis further shows wave acceleration in close proximity of the spacer with subsequent deceleration further downstream. Analyzing the wave as a two-dimensional entity it can be seen that the wave paths are clearly structured by the spacer and hence do not travel circumferentially around the fuel rod. Wave separation and merge has a

  2. Experimental studies of the effect of functional spacers to annular flow in subchannels of a BWR fuel element

    International Nuclear Information System (INIS)

    Damsohn, M.; Prasser, H.-M.

    2010-01-01

    For the prediction of dryout in fuel elements of boiling water reactors, the dynamic behavior of the water film covering the fuel rod has to be understood. This paper provides high resolved experimental data of the liquid film and gives insight into the dynamic film behavior. The experiments of this work were conducted in a vertical channel representing a pair of neighboring subchannels of a BWR fuel rod bundle. Air and water at ambient pressure and temperature are used as model fluids, creating an annular flow in the test section. The influence of different functional spacer shapes on the liquid film has been studied. The heart of the instrumentation is a liquid film sensor (LFS), which measures the film thickness distribution around a half cylinder with a matrix of 64 x 16 measuring points with a time resolution of 10,000 frames per second and a spatial resolution of 2 mm x 2 mm. The high resolution allows for a visualization of the dynamic liquid film as a movie animation. Principals of the dynamic behavior of the liquid film are observed. The time-averaged film thickness distributions show that the spacers structure the liquid film significantly. The gaseous phase is accelerated due to the cross-section blockage caused by the spacer. This leads to a local thinning of the liquid film downstream of the spacer. Two statistical evaluation methods are presented to determine different dynamic wave properties: The wave velocity as function of the wave height, the traveling path of the waves and the location of wave separation and merge events. The first evaluation method shows that big waves move generally faster than small waves. The analysis further shows wave acceleration in close proximity of the spacer with subsequent deceleration further downstream. Analyzing the wave as a two-dimensional entity it can be seen that the wave paths are clearly structured by the spacer and hence do not travel circumferentially around the fuel rod. Wave separation and merge has a

  3. Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

    Science.gov (United States)

    Martínez-Torrents, Albert; Serrano-Purroy, Daniel; Sureda, Rosa; Casas, Ignasi; de Pablo, Joan

    2017-05-01

    The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

  4. Instant release fraction corrosion studies of commercial UO{sub 2} BWR spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martínez-Torrents, Albert, E-mail: albert.martinez@ctm.com.es [Fundació CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Serrano-Purroy, Daniel [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, P.O. Box 2340, D-76125 Karlsruhe (Germany); Sureda, Rosa [Fundació CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Casas, Ignasi [Department of Chemical Engineering, Universitat Politècnica de Catalunya – Barcelona Tech, Eduard Maristany 14, 08019 Barcelona (Spain); Pablo, Joan de [Fundació CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Department of Chemical Engineering, Universitat Politècnica de Catalunya – Barcelona Tech, Eduard Maristany 14, 08019 Barcelona (Spain)

    2017-05-15

    The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

  5. Study of check valve slamming in a BWR feedwater system following a postulated pipe break

    International Nuclear Information System (INIS)

    Safwat, H.H.; Arastu, A.H.; Norman, A.

    1985-01-01

    This study deals with a swing check valve slamming due to a break at relatively short distance from the valve. Under this situation, substantial flashing occurs near the valve and the result of the study are subject to what is believed to be a conservative simplifying assumption, i.e., the hydrodynamic moment acting on the valve during the transient is represented by resultant moment due to the pressure differential across the valve. It is believed that vapor voids forming at the valve would actually reduce the disk impact velocities in comparison to those predicted under this simplifying assumption. A technique used to represent a double-ended break through hypothetical valves may have some influence on the results particularly for long break opening times. The study has yielded good insight to help understand the complex problem. The study has focused on some parameters and the reader may raise questions on the effects of other parameters. Nevertheless, the present study underlines the complexity facing analysts dealing with this transient using analytical methods. Though some experimental data are available, the authors believe that an experimental study (recognizing the complexity of the experimental setup and instrumentation), would be quite useful. It can provide answers to questions facing analysts dealing with this problem and thus avoid unnecessary conservatisms due to uncertainties in input data

  6. 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)

  7. Theoretical and Experimental Studies of Magneto-Rayleigh-Taylor Instabilities

    International Nuclear Information System (INIS)

    Lau, Yue Ying; Gilgenbach, Ronald

    2013-01-01

    Magneto-Rayleigh-Taylor instability (MRT) is important to magnetized target fusion, wire-array z-pinches, and equation-of-state studies using flyer plates or isentropic compression. It is also important to the study of the crab nebula. The investigators performed MRT experiments on thin foils, driven by the mega-ampere linear transformer driver (LTD) facility completed in their laboratory. This is the first 1-MA LTD in the USA. Initial experiments on the seeding of MRT were performed. Also completed was an analytic study of MRT for a finite plasma slab with arbitrary magnetic fields tangential to the interfaces. The effects of magnetic shear and feedthrough were analyzed

  8. Theoretical and Experimental Studies of Magneto-Rayleigh-Taylor Instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Lau, Yue Ying [University of Michigan, Ann Arbor, MI (United States); Gilgenbach, Ronald [University of Michigan, Ann Arbor, MI (United States)

    2013-07-07

    Magneto-Rayleigh-Taylor instability (MRT) is important to magnetized target fusion, wire-array z-pinches, and equation-of-state studies using flyer plates or isentropic compression. It is also important to the study of the crab nebula. The investigators performed MRT experiments on thin foils, driven by the mega-ampere linear transformer driver (LTD) facility completed in their laboratory. This is the first 1-MA LTD in the USA. Initial experiments on the seeding of MRT were performed. Also completed was an analytic study of MRT for a finite plasma slab with arbitrary magnetic fields tangential to the interfaces. The effects of magnetic shear and feedthrough were analyzed.

  9. Sensitivity study on heuristic rules applied to the neutronic optimization of cells for BWR

    International Nuclear Information System (INIS)

    Gonzalez C, J.; Martin del Campo M, C.; Francois L, J.L.

    2004-01-01

    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)

  10. A study on the boron injection initiation temperature curve of BWR

    International Nuclear Information System (INIS)

    Wang, S.-J.; Chien, C.-S.; Fann, S.-Y.; Chiang, S.-C.

    2007-01-01

    Boron injection initiation temperature (BIIT) provides important information for the safe shutdown of the reactor using boron injection system during anticipated transient without scram (ATWS). The purpose of this paper is to study BIIT curve of boiling water reactor owners' group (BWROG). The unreasonable and non-conservative parts of BIIT are pointed out and suggested modifications are made. The starting reactor power of BIIT is increased in order to meet the actual application. The lower limit of suppression pool temperature of BIIT is revised for conservative operation during ATWS conditions. Analysis of the effects of maximum temperature capacity of the suppression chamber and concentration of boron in standby liquid control tank shows that BIIT is decreased by adopting a more conservative value of maximum temperature capacity of the suppression chamber. Consequently, early boron injection is anticipated. For system with automatic boron injection system, BIIT is not required

  11. TEM/STEM study of Zircaloy-2 with protective FeAl(Cr) layers under simulated BWR environment and high-temperature steam exposure

    Science.gov (United States)

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

    2018-04-01

    FeAl(Cr) thin-film depositions on Zircaloy-2 were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) with respect to oxidation behavior under simulated boiling water reactor (BWR) conditions and high-temperature steam. Columnar grains of FeAl with Cr in solid solution were formed on Zircaloy-2 coupons using magnetron sputtering. NiFe2O4 precipitates on the surface of the FeAl(Cr) coatings were observed after the sample was exposed to the simulated BWR environment. High-temperature steam exposure resulted in grain growth and consumption of the FeAl(Cr) layer, but no delamination at the interface. Outward Al diffusion from the FeAl(Cr) layer occurred during high-temperature steam exposure (700 °C for 3.6 h) to form a 100-nm-thick alumina oxide layer, which was effective in mitigating oxidation of the Zircaloy-2 coupons. Zr intermetallic precipitates formed near the FeAl(Cr) layer due to the inward diffusion of Fe and Al. The counterflow of vacancies in response to the Al and Fe diffusion led to porosity within the FeAl(Cr) layer.

  12. Residual stress analysis in BWR pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Leung, C.P.; Pont, D.

    1992-06-01

    Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

  13. Moderator temperature coefficient in BWR core

    International Nuclear Information System (INIS)

    Naito, Yoshitaka

    1977-01-01

    Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

  14. IGSCC crack propagation rate measurement in BWR environments. Executive summary of a Round Robin study

    International Nuclear Information System (INIS)

    Andresen, Peter L.

    1998-01-01

    Five of the world's best laboratories at performing stress corrosion crack growth studies - ABB Atom AB, AEA Technology, GE Corporate Research and Development Center, Studsvik Material AB, and VTT Manufacturing Technology, were selected to participate in a round robin to evaluate the quality and reproducibility of testing conditions and resulting stress corrosion crack growth rates in sensitized type 304 stainless steel in 288 deg C water. Heat treated, machined and fatigue pre-cracked specimens were provided to all laboratories, and detailed test procedures prescribed the use of active loading, reversed dc potential drop crack monitoring, a common reference electrode supplied to all laboratories by GE CRD (to be used along side each laboratory's own reference electrode), and highly specified water chemistry conditions. The ability of each laboratory to achieve optimal testing conditions varied, although all laboratories achieved an impressive standard of testing control. The most significant laboratory-to-laboratory differences were associated with their ability to achieve high purity autoclave outlet water, reproduce accurate measurements of corrosion potential on the test specimen, and provide high resolution crack following using a reversed dc potential drop. However, the most notable outcome of the program is the consistent observation by all laboratories that initiating and sustaining stress corrosion crack growth at constant load in sensitized type 304 stainless steel is difficult, despite the use of a moderately high stress intensity, and high dissolved oxygen and corrosion potential conditions. Concerns for specimen machining and pre-cracking were identified, although these factors were not the sole cause of difficulty in initiating and sustaining stress corrosion cracking. It was shown that many phases of specimen preparation and testing can have a large influence on the measured SCC response. Even under the best test conditions it is critical to ensure

  15. IGSCC crack propagation rate measurement in BWR environments. Executive summary of a Round Robin study

    Energy Technology Data Exchange (ETDEWEB)

    Andresen, Peter L. [GE Corporate Research and Development, Schenectady, NY (United States)

    1998-12-31

    Five of the world`s best laboratories at performing stress corrosion crack growth studies - ABB Atom AB, AEA Technology, GE Corporate Research and Development Center, Studsvik Material AB, and VTT Manufacturing Technology, were selected to participate in a round robin to evaluate the quality and reproducibility of testing conditions and resulting stress corrosion crack growth rates in sensitized type 304 stainless steel in 288 deg C water. Heat treated, machined and fatigue pre-cracked specimens were provided to all laboratories, and detailed test procedures prescribed the use of active loading, reversed dc potential drop crack monitoring, a common reference electrode supplied to all laboratories by GE CRD (to be used along side each laboratory`s own reference electrode), and highly specified water chemistry conditions. The ability of each laboratory to achieve optimal testing conditions varied, although all laboratories achieved an impressive standard of testing control. The most significant laboratory-to-laboratory differences were associated with their ability to achieve high purity autoclave outlet water, reproduce accurate measurements of corrosion potential on the test specimen, and provide high resolution crack following using a reversed dc potential drop. However, the most notable outcome of the program is the consistent observation by all laboratories that initiating and sustaining stress corrosion crack growth at constant load in sensitized type 304 stainless steel is difficult, despite the use of a moderately high stress intensity, and high dissolved oxygen and corrosion potential conditions. Concerns for specimen machining and pre-cracking were identified, although these factors were not the sole cause of difficulty in initiating and sustaining stress corrosion cracking. It was shown that many phases of specimen preparation and testing can have a large influence on the measured SCC response. Even under the best test conditions it is critical to ensure

  16. Numerical study of two-dimensional moist symmetric instability

    Directory of Open Access Journals (Sweden)

    M. Fantini

    2008-06-01

    Full Text Available The 2-D version of the non-hydrostatic fully compressible model MOLOCH developed at ISAC-CNR was used in idealized set-up to study the start-up and finite amplitude evolution of symmetric instability. The unstable basic state was designed by numerical integration of the equation which defines saturated equivalent potential vorticity qe*. We present the structure and growth rates of the linear modes both for a supersaturated initial state ("super"-linear mode and for a saturated one ("pseudo"-linear mode and the modifications induced on the base state by their finite amplitude evolution.

  17. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  18. NUMERICAL STUDY OF THE VISHNIAC INSTABILITY IN SUPERNOVA REMNANTS

    International Nuclear Information System (INIS)

    Michaut, C.; Cavet, C.; Bouquet, S. E.; Roy, F.; Nguyen, H. C.

    2012-01-01

    The Vishniac instability is thought to explain the complex structure of radiative supernova remnants in their Pressure-Driven Thin Shell (PDTS) phase after a blast wave (BW) has propagated from a central explosion. In this paper, the propagation of the BW and the evolution of the PDTS stage are studied numerically with the two-dimensional (2D) code HYDRO-MUSCL for a finite-thickness shell expanding in the interstellar medium (ISM). Special attention is paid to the adiabatic index, γ, and three distinct values are taken for the cavity (γ 1 ), the shell (γ 2 ), and the ISM (γ 3 ) with the condition γ 2 1 , γ 3 . This low value of γ 2 accounts for the high density in the shell achieved by a strong radiative cooling. Once the spherical background flow is obtained, the evolution of a 2D-axisymmetric perturbation is computed from the linear to the nonlinear regime. The overstable mechanism, previously demonstrated theoretically by E. T. Vishniac in 1983, is recovered numerically in the linear stage and is expected to produce and enhance anisotropies and clumps on the shock front, leading to the disruption of the shell in the nonlinear phase. The period of the increasing oscillations and the growth rate of the instability are derived from several points of view (the position of the perturbed shock front, mass fluxes along the shell, and density maps), and the most unstable mode differing from the value given by Vishniac is computed. In addition, the influence of several parameters (the Mach number, amplitude and wavelength of the perturbation, and adiabatic index) is examined and for wavelengths that are large enough compared to the shell thickness, the same conclusion arises: in the late stage of the evolution of the radiative supernova remnant, the instability is dampened and the angular initial deformation of the shock front is smoothed while the mass density becomes uniform with the angle. As a result, our model shows that the supernova remnant returns to a

  19. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  20. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

  1. The noise analysis and the BWR operation map

    International Nuclear Information System (INIS)

    Blazquez, J.; Ballestrin, J.

    1996-01-01

    An analytical expression for the Decay Ratio is obtained: DR = exp(-bW / P 1/2 ). The physics behind is also explained. It applies to a commercial BWR Operation Map, on the vicinity of the power instability. This functional form seems fitting to the structure of the Operation map. The power P and the coolant flow are measured straightforward; the Decay Ratio is obtained by neutron noise analysis techniques. The parameter b, depending on the void reactivity coefficient, is then calculated on line during the Reactor Operation. New DR value is now predicted for each new displacement on the Map, so unexpected instability events are more likely avoided. (authors)

  2. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  3. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  4. 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)

  5. BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    Anstine, L.D.

    1983-05-01

    This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

  6. Studies of Corrosion of Cladding Materials in Simulated BWR-environment Using Impedance Measurements. Part I: Measurements in the Pre-transition Region

    International Nuclear Information System (INIS)

    Forsberg, Stefan; Ahlberg, Elisabet; Andersson, Ulf

    2004-09-01

    The corrosion of three Zircaloy 2 cladding materials, LK2, LK2+ and LK3, have been studied in-situ in an autoclave using electrochemical impedance spectroscopy. Measurements were performed in simulated BWR water at temperatures up to 288 deg C. The impedance spectra were successfully modelled using equivalent circuits. When the oxide grew thicker during the experiments, a change-over from one to two time constants was seen, showing that a layered structure was formed. Oxide thickness, oxide conductivity and effective donor density were evaluated from the impedance data. The calculated oxide thickness at the end of the experiments was consistent with the value obtained from SEM. It was shown that the difference in oxide growth rate between the investigated materials is small in the pre-transition region. The effective donor density, which is a measure of electronic conductivity, was found to be lower for the LK3 material compared to the other two materials

  7. Longitudinal coupled-bunch instability studies in the PS

    CERN Document Server

    Damerau, H

    2017-01-01

    The main longitudinal limitation for LHC-type beams inthe PS are coupled-bunch instabilities. A dedicated proto-typefeedbacksystemusingaFinemetcavityasalongitudinalkicker has been installed. Extensive tests with beam havebeen performed to explore the intensity reach with this feed-back. The maximum intensity with nominal longitudinalemittance at PS extraction has been measured, as well as theemittance required to keep the beam longitudinally stableat the design intensity for the High-Luminosity LHC (HL-LHC). A higher-harmonic cavity is a complementary op-tion to extend the intensity reach beyond the capabilities ofthe coupled-bunch feedback. Preliminary machine develop-ment (MD) studies operating one20MHzor one40MHzRF system as a higher harmonic at the flat-top indicate thebeneficial effect on longitudinal beam stability

  8. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  9. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  10. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  11. Familial colorectal cancer, can it be identified by microsatellite instability and chromosomal instability? - A case-control study

    DEFF Research Database (Denmark)

    Sunde, Lone; Bisgaard, Marie Luise; Soll-Johanning, Helle

    2009-01-01

    (Chromosome INstability=LOH (loss of heterozygosity) and/or DNA-aneuploidy (abnormal nuclear DNA contents)) could be used as predictors of familial CRC. Formalin-fixed tissue from 97 patients with CRC (29 patients with 2 or more affected first-degree relatives (="cases"), 29 matched CRC controls without......Colonoscopy is recommended for persons with a familial risk of colorectal cancer (CRC). A familial risk is identified by a family history with CRC and/or predisposing mutation(s). However, such information may not be available. We analysed whether MSI (MicroSatellite Instability) and/or CIN...... a family history, and 39 relatives to cases) were analysed for MSI and CIN. In this small case-control study, no significant differences in the frequencies of MSI and CIN were observed between cases with a family history and their controls without a family history. MSI+;CIN- was observed in 6/29 cases...

  12. Studies of the longitudinal instability with an electron beam

    International Nuclear Information System (INIS)

    1993-01-01

    Goals for our first-year period are as follows: To study the evolution of a small perturbation in the current pulse (introduced via the grid voltage on the electron gun) when the beam propagates through our 5-m long periodic solenoid channel. Specifically, to see if the perturbation is reflected from the rear end of the pulse. So far these objectives have been met without any delays. We were able to launch different perturbations on the beam resulting in either a slow space-charge wave or a fast wave or both waves. The relative strength of each wave was found to depend on the electron emission temperature of the cathode. The propagation of these waves on an initially rectangular longitudinal beam profile was measured with fast current monitors and the kinetic energy was measured with sensitive energy analyzers at various positions along the 5-m long solenoidal focusing channel. We have also begun to study the behavior of the waves when they reach the respective edge of the beam. But this work is still of a preliminary nature, and we need to refine the beam conditions and measurements in future studies to reach any firm conclusions. Preparations for the resistive-wall instability experiment are in progress

  13. Experimental study of flow instability and CHF in a natural circulation system with subcooled boiling

    International Nuclear Information System (INIS)

    Yang, R.C.; Shi, D.Q.; Lu, Z.Q.; Zheng, R.C.; Wang, Y.

    1996-01-01

    Experimental study has been performed to investigate flow instability and critical heat flux (CHF) in a natural circulation system with subcooled boiling. In the experiments three kinds of heated sections were used. Freon-12 was used as the working medium. The experiments show which one of the two phenomena, flow instability and CHF condition, may first occur in the system depends on not only the heat input power to the heated section and the parameters of the working medium, but also the construction of the heated section. The occurrence of the flow instability mainly depends on the total heat input power to the heated section and the CHF condition is mainly caused by the local heat flux of the heated section. In the experiments two kinds of flow instability, flow instability with high frequency and flow instability with low frequency, were found. But they all belong to density wave instability. The influence of the parameters of the working medium on the onset of the flow instability and CHF condition in the system were investigated. The stability boundaries were determined through the experiments. By means of dimensional analysis of integral equations, a common correlation describing the threshold condition of onset of the flow instability was obtained

  14. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  15. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  16. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  17. LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

    2005-01-01

    Full text of publication follows: Unstable oscillation of a nuclear power reactor core is one of the main reasons that causes minor core damage. Stability analysis needs to be performed to predict the potential problem as early as possible and to prevent core instability events from happening. Nuclear Regulatory Commission (NRC) requests all BWR licensees to examine each core reload and to impose operating limitations, as appropriate, to ensure compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. Therefore, the core instability is directly related to the fuel design limits. The core and channel DR (decay ratio) calculation are commonly performed to determine system's stability when new fuel designs are introduced in the core. In order to establish the independent analysis technology for BWR licensees and verifications, the Institute of Nuclear Energy Research (INER) has obtained agreement from NRC and implemented the 'Methodology and Procedure for Calculation of Core and Channel Decay Ratios with LAPUR', which was developed by the IBERINCO in 2001. LAPUR5 uses a multi-nodal description of the neutron dynamics, together with a distributed parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations around a steady state condition. From the output of LAPUR5, the following results are obtained: global core decay ratio, out-of phase core decay ratio, and channel decay ratio. They are key parameters in the determination of BWR core stability

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

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

    2011-01-01

    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

  19. Study on flow instabilities in two-phase mixtures

    International Nuclear Information System (INIS)

    Ishii, M.

    1976-03-01

    Various mechanisms that can induce flow instabilities in two-phase flow systems are reviewed and their relative importance discussed. In view of their practical importance, the density-wave instabilities have been analyzed in detail based on the one-dimensional two-phase flow formulation. The dynamic response of the system to the inlet flow perturbations has been derived from the model; thus the characteristic equation that predicts the onset of instabilities has been obtained. The effects of various system parameters, such as the heat flux, subcooling, pressure, inlet velocity, inlet orificing, and exit orificing on the stability boundary have been analyzed. In addition to numerical solutions, some simple stability criteria under particular conditions have been obtained. Both results have been compared with various experimental data, and a satisfactory agreement has been demonstrated

  20. Identification of the reduced order models of a BWR reactor

    International Nuclear Information System (INIS)

    Hernandez S, A.

    2004-01-01

    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)

  1. Experimental study of parametric instabilities and anomalous heating in plasma

    International Nuclear Information System (INIS)

    Batanov, G.M.; Rabinovich, M.S.

    1975-01-01

    Over the last few years the study of the dissipation of electromagnetic wave energy in a hot plasma has become perhaps one of the main problems of high-temperature plasma physics and controlled thermonuclear fusion. The focus of attention is on the processes by which electromagnetic energy is transformed into potential plasma waves and the processes involving relaxation of the latter. In this paper the authors summarize the experimental research into these processes conducted at the Lebedev Physics Institute over the 10 cm wave band. In the case of an isotropic plasma the authors recorded non-linear generation of Langmuir noise, the energy density of which was found to be comparable, in order of magnitude, with that of a pump wave. They detected the generation of fast-electron streams, the non-stationary character of the latter with respect to time, and non-linear transmissivity of the plasma layer. In the case of a magnetoactive plasma they studied the parametric excitation of oscillations at the upper hybrid frequency during its resonance with the first overtone of the pump wave. Excitation of plasma noise was found to be accompanied by a flux of fast-electrons, in the energy spectrum of which separate groups were detected. It was also found that the effective collision frequency increased by 1-3 orders, compared to the pari-collision frequency. In the region of magnetic waves close to the electron cyclotron resonance the authors observed forced Mandel'shtam-Brillouin scattering and kinetic instability of the plasma. It was found that the excitation of ionic Langmuir noise preceded ''anomalous absorption'' of waves and ''anomalous heating'' of electrons. The authors further consider the possibility of an experimental study of anomalous heating in plasma in the region of the lower hybrid frequencies, using the Institute's L-2 stellarator. (author)

  2. BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

    International Nuclear Information System (INIS)

    Pohl, P.; Wehle, F.

    2010-01-01

    Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

  3. Study of surface and bulk instabilities in MHD duct flow with imitation of insulator coating imperfections

    Energy Technology Data Exchange (ETDEWEB)

    Xu Zengyu [Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan 610041 (China)]. E-mail: xuzy@swip.ac.cn; Pan Chuanjie [Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan 610041 (China); Wei Wenhao [Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan 610041 (China); Kang Weishan [Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan 610041 (China)

    2006-02-15

    MHD phenomena in a duct flow were studied experimentally by using copper electrodes inserted into the wall of a perfectly insulated duct. The electrodes were connected using a copper wire to imitate different insulator coating imperfection conditions. The experimental results show instabilities of electric potential at the wall (surface instabilities) as well as instabilities in the pressure and velocity (bulk instabilities). The instabilities are strongly dependent on the scale of the copper wire. Three different cases were studied (at the same flow regimes, but with different electrode connections), where the potential at the duct wall is smaller, equal to or higher than the product of duct diameter 2a and transverse magnetic field B and average velocity V . MHD pressure drop {delta}P also exhibits significant changes.

  4. Numerical Studies of Electromagnetic Instabilities in Intense Charged Particle Beams with Large Energy Anisotropy

    CERN Document Server

    Startsev, Edward; Lee, Wei-li

    2005-01-01

    In intense charged particle beams with large energy anisotropy, free energy is available to drive transverse electromagnetic Weibel-type instabilities. Such slow-wave transverse electromagnetic instabilities can be described by the so-called Darwin model, which neglects the fast-wave portion of the displacement current. The Weibel instability may also lead to an increase in the longitudinal velocity spread, which would make the focusing of the beam difficult and impose a limit on the minimum spot size achievable in heavy ion fusion experiments. This paper reports the results of recent numerical studies of the Weibel instability using the Beam Eigenmode And Spectra (bEASt) code for space-charge-dominated, low-emittance beams with large tune depression. To study the nonlinear stage of the instability, the Darwin model is being developed and incorporated into the Beam Equilibrium Stability and Transport(BEST) code.

  5. The use of the partial coherence function technique for the investigation of BWR noise dynamics

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1983-01-01

    The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

  6. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

  7. Instability study during implosion in the Tupa Theta-Pinch

    International Nuclear Information System (INIS)

    Kayama, M.E.; Boeckelmann, H.K.

    1986-01-01

    The importance of instabilities which occur during plasma heating in a Theta Pinch, in the implosion phase, is analysed. The plasma diagnostic was done by ultrafast photography and diamagnetic probe. The implosion time and the current layer thickness were calculated using a hybrid code for plasma simulation. The theoretical data were compared with the experimental ones. (M.C.K.) [pt

  8. 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)

  9. Analysis of the Cofrentes instability with the Hilbert-Huang transform

    International Nuclear Information System (INIS)

    Blazquez, J.; Galindo, A.

    2010-01-01

    The most obvious application of the Hilbert-Huang transform is the denoising (signal isolation). In this article, the dynamic system is the power of a BWR reactor that undergoes instability. The signal and the dynamic systems are described, which in this case corresponds to a current incident in a commercial BWR reactor (Cofrentes). Finally, empirical modes are calculated and the results are analyzed.

  10. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  11. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  12. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  13. Experimental study of natural circulation flow instability in rectangular channels

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Tao; Qi, Shi; Song, Mingqiang [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Passive Nuclear Safety Technology, Beijing (China). Beijing Key Lab.; Xiao, Zejun [Nuclear, Reactor Thermal Hydraulics Technology, Chengdu (China). CNNC Key Lab.

    2017-05-15

    Experiments of natural circulation flow instability were conducted in rectangular channels with 5 mm and 10 mm wide gaps. Results for different heating powers were obtained. The results showed that the flow will tend to be instable with the growing of heating power. The oscillation period of pressure D-value and volume flow are the same, but their phase positions are opposite. They both can be described by trigonometric functions. The existence of edge position and secondary flow will strengthen the disturbance of fluid flow in rectangle channels, which contributes to heat transfer. The disturbance of bubble and fluid will be strengthened, especially in the saturated boiling section, which make it possible for the mixing flow. The results also showed that the resistance in 5 mm channel is bigger than that in 10 mm channel, it is less likely to form stable natural circulation in the subcooled region.

  14. Local study of helical magnetorotational instability in viscous Keplerian disks

    Science.gov (United States)

    MahdaviGharavi, M.; Hajisharifi, K.; Mehidan, H.

    2018-03-01

    In this paper, regarding the recent detection of significant azimuthal magnetic field in some accretion disks such as protostellar (Donati et al. in Nature 438:466, 2005), the multi-fluid model has been employed to analysis the stability of Keplerian rotational viscous dusty plasma system in a current-free helical magnetic field structure. Using the fluid-Maxwell equations, the general dispersion relation of the excited modes in the system has been obtained by applying the local approximation method in the linear perturbation theory. The typical numerical analysis of the obtained dispersion relation in the high-frequency regime shows that the presence of azimuthal magnetic field component in Keplerian flow has a considerable role in the stability conditions of the system. It also shows that the magnetic field helicity has a stabilization role against the magnetorotational instability (MRI) in the system due to contraction of the unstable wavelength region and decreasing the maximum growth rate of the instability. In this sense, the stabilization role of the viscosity term is more considerable for HMRI (instability in the presence of azimuthal magnetic field component) than the corresponding MRI (instability in the absence of azimuthal magnetic field component). Moreover, considering the discovered azimuthal magnetic field in these systems, the MRI can be arisen in the over-all range of dust grains construction values in contract with traditional MRI. This investigation can greatly contribute to better understanding the physics of some astrophysical phenomena, such as the main source of turbulence and angular momentum transport in protostellar and the other sufficiently ionized astrophysical disks, where the azimuthal magnetic field component in these systems can play a significant role.

  15. Study on Influence of Tube Arrays on Fluid Elastic Instability

    Science.gov (United States)

    Ishihara, Kunihiko; Kitayama, Gen

    The tube bank is used in boilers, heat exchangers in power plants and steam generators in nuclear plants. These tubes sometimes vibrate violently and come to the fatigue failure due to the flow induced vibration which is caused by the cross flow. This phenomenon is that the large vibrations arise at the critical flow velocity and it is called fluid elastic instability. However the relation between the onset velocity of fluid elastic instability and the tube array's geometry has not been clarified sufficiently. There is a few reference related to the relation between the pitch to diameter ratio and the onset velocity even in the lattice arrays. In this paper, the influence of tube arrays on fluid elastic instability is examined by experiments. As a result, it is clarified that the tube vibrations become large as T/D increases and L/D decreases, and the tube vibrations strongly depend on the dynamic characteristics of tubes such as the natural frequency and the damping ability.

  16. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

    2006-01-01

    This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

  17. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  18. A numerical study of Richtmyer endash Meshkov instability driven by cylindrical shocks

    International Nuclear Information System (INIS)

    Zhang, Q.; Graham, M.J.

    1998-01-01

    As an incident shock wave hits a material interface between two fluids of different densities, the interface becomes unstable. Small disturbances at the interface start to grow. This interfacial instability is known as a Richtmyer endash Meshkov (RM) instability. It plays an important role in the studies of inertial confinement fusion and supernova. The majority of studies of the RM instability were in plane geometry emdash namely, plane shocks in Cartesian coordinates. We present a systematic numerical study of the RM instability driven by cylindrical shocks for both the imploding and exploding cases. The imploding (exploding) case refers to a cylindrical shock colliding with the material interface from the outside in (inside out). The phenomenon of reshock caused by the waves reflected from the origin is also studied. A qualitative understanding of this system has been achieved. Detailed studies of the growth rate of the fingers at the unstable interface are presented. copyright 1998 American Institute of Physics

  19. High pressure and synchrotron radiation studies of solid state electronic instabilities

    International Nuclear Information System (INIS)

    Pifer, J.H.; Croft, M.C.

    1992-04-01

    This report discusses Eu and General Valence Instabilities; Ce Problem: L 3 Spectroscopy Emphasis; Bulk Property Emphasis; Transition Metal Compound Electronic Structure; Electronic Structure-Phonon Coupling Studies; High Temperature Superconductivity and Oxide Materials; and Novel Materials Collaboration with Chemistry

  20. A simplified spatial model for BWR stability

    International Nuclear Information System (INIS)

    Berman, Y.; Lederer, Y.; Meron, E.

    2012-01-01

    A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

  1. Comparative Study of Magnetic Instabilities in Cerium Compounds

    International Nuclear Information System (INIS)

    Pedrazzini, Pablo

    2003-01-01

    The variety of new phases and physical phenomena discovered in intermetallic compounds containing Rare Earths or Actinides has motivated, during the last four decades, the sustained study of their magnetic phase diagrams.The current interest is focused on the investigation of the region of the phase diagram where the magnetic order of Cerium, Ytterbium and Uranium based systems is destabilized.In this region different behaviours have been detected, such as non conventional superconductivity and the anomalous dependencies of the thermal, magnetic and transport properties at very low temperatures, associated to non-Fermi liquid behaviour.A simple model, the Doniach diagram, has guided the interpretation of the destabilization of the magnetic order in the previously mentioned systems.However, most of the systems that have been studied so far cannot be described within this model.This fact has motivated the development of a phenomenological classification of phase diagrams that has been mostly applied to cerium based compounds.This classification defines three types of phase diagrams, that can be distinguished by the way in which the magnetic transition is suppressed when a control parameter (such as doping or pressure) is driven towards its critical value.Within this scenario, we study the suppression of the antiferromagnetic order of the intermetallic compounds CeIn 3 , CeRh 2 Si 2 and CePd 2 Al 3 as a function of Ce-ligand alloying.The resulting systems, CeIn 3-x Sn x , Ce(Cu x Rh 1-x ) 2 Si 2 and CePd 2-x Ni x Al 3 , present different crystalline structures and the effects produced by the alloying process are different in each case.We analyse the resulting magnetic phase diagrams, and compare them with the above mentioned phenomenological classification.With such a purpose, we study in detail the region in which the magnetic instability takes place, in the proximity of the respective critical concentrations.Taking into account both our results and those reported in

  2. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

  3. Laboratory Study of Magnetorotational Instability and Hydrodynamic Stability at Large Reynolds Numbers

    Science.gov (United States)

    Ji, H.; Burin, M.; Schartman, E.; Goodman, J.; Liu, W.

    2006-01-01

    Two plausible mechanisms have been proposed to explain rapid angular momentum transport during accretion processes in astrophysical disks: nonlinear hydrodynamic instabilities and magnetorotational instability (MRI). A laboratory experiment in a short Taylor-Couette flow geometry has been constructed in Princeton to study both mechanisms, with novel features for better controls of the boundary-driven secondary flows (Ekman circulation). Initial results on hydrodynamic stability have shown negligible angular momentum transport in Keplerian-like flows with Reynolds numbers approaching one million, casting strong doubt on the viability of nonlinear hydrodynamic instability as a source for accretion disk turbulence.

  4. Quantitative study of the trapped particle bunching instability in Langmuir waves

    International Nuclear Information System (INIS)

    Hara, Kentaro; Boyd, Iain D.; Chapman, Thomas; Joseph, Ilon; Berger, Richard L.; Banks, Jeffrey W.; Brunner, Stephan

    2015-01-01

    The bunching instability of particles trapped in Langmuir waves is studied using Vlasov simulations. A measure of particle bunching is defined and used to extract the growth rate from numerical simulations, which are compared with theory [Dodin et al., Phys. Rev. Lett. 110, 215006 (2013)]. In addition, the general theory of trapped particle instability in 1D is revisited and a more accurate description of the dispersion relation is obtained. Excellent agreement between numerical and theoretical predictions of growth rates of the bunching instability is shown over a range of parameters

  5. Quantitative study of the trapped particle bunching instability in Langmuir waves

    Energy Technology Data Exchange (ETDEWEB)

    Hara, Kentaro, E-mail: kenhara@umich.edu; Boyd, Iain D. [Department of Aerospace Engineering, University of Michigan, Ann Arbor, Michigan 48109 (United States); Chapman, Thomas; Joseph, Ilon; Berger, Richard L. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Banks, Jeffrey W. [Department of Mathematical Sciences, Rensselaer Polytechnic Institute, Troy, New York 12180 (United States); Brunner, Stephan [Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, Ecole Polytechnique Fédérale de Lausanne, CRPP-PPB, CH-1015 Lausanne (Switzerland)

    2015-02-15

    The bunching instability of particles trapped in Langmuir waves is studied using Vlasov simulations. A measure of particle bunching is defined and used to extract the growth rate from numerical simulations, which are compared with theory [Dodin et al., Phys. Rev. Lett. 110, 215006 (2013)]. In addition, the general theory of trapped particle instability in 1D is revisited and a more accurate description of the dispersion relation is obtained. Excellent agreement between numerical and theoretical predictions of growth rates of the bunching instability is shown over a range of parameters.

  6. Effect of nonlinear void reactivity on bifurcation characteristics of a lumped-parameter model of a BWR: A study relevant to RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2017-04-15

    Highlights: • A simplified model with nonlinear void reactivity feedback is studied. • Method of multiple scales for nonlinear analysis and oscillation characteristics. • Second order void reactivity dominates in determining system dynamics. • Opposing signs of linear and quadratic void reactivity enhances global safety. - Abstract: In the present work, the effect of nonlinear void reactivity on the dynamics of a simplified lumped-parameter model for a boiling water reactor (BWR) is investigated. A mathematical model of five differential equations comprising of neutronics and thermal-hydraulics encompassing the nonlinearities associated with both the reactivity feedbacks and the heat transfer process has been used. To this end, we have considered parameters relevant to RBMK for which the void reactivity is known to be nonlinear. A nonlinear analysis of the model exploiting the method of multiple time scales (MMTS) predicts the occurrence of the two types of Hopf bifurcation, namely subcritical and supercritical, leading to the evolution of limit cycles for a range of parameters. Numerical simulations have been performed to verify the analytical results obtained by MMTS. The study shows that the nonlinear reactivity has a significant influence on the system dynamics. A parametric study with varying nominal reactor power and operating conditions in coolant channel has also been performed which shows the effect of change in concerned parameter on the boundary between regions of sub- and super-critical Hopf bifurcations in the space constituted by the two coefficients of reactivities viz. the void and the Doppler coefficient of reactivities. In particular, we find that introduction of a negative quadratic term in the void reactivity feedback significantly increases the supercritical region and dominates in determining the system dynamics.

  7. Nuclear-coupled thermal-hydraulic nonlinear stability analysis using a novel BWR reduced order model. Pt. 1. The effects of using drift flux versus homogeneous equilibrium models

    International Nuclear Information System (INIS)

    Dokhane, A.; Henning, D.; Chawla, R.; Rizwan-Uddin

    2003-01-01

    BWR stability analysis at PSI, as at other research centres, is usually carried out employing complex system codes. However, these do not allow a detailed investigation of the complete manifold of all possible solutions of the associated nonlinear differential equation set. A novel analytical, reduced order model for BWR stability has been developed at PSI, in several successive steps. In the first step, the thermal-hydraulic model was used for studying the thermal-hydraulic instabilities. A study was then conducted of the one-channel nuclear-coupled thermal-hydraulic dynamics in a BWR by adding a simple point kinetic model for neutron kinetics and a model for the fuel heat conduction dynamics. In this paper, a two-channel nuclear-coupled thermal-hydraulic model is introduced to simulate the out-of phase oscillations in a BWR. This model comprises three parts: spatial mode neutron kinetics with the fundamental and fist azimuthal modes; fuel heat conduction dynamics; and thermal-hydraulics model. This present model is an extension of the Karve et al. model i.e., a drift flux model is used instead of the homogeneous equilibrium model for two-phase flow, and lambda modes are used instead of the omega modes for the neutron kinetics. This two-channel model is employed in stability and bifurcation analyses, carried out using the bifurcation code BIFDD. The stability boundary (SB) and the nature of the Poincare-Andronov-Hopf bifurcation (PAF-B) are determined and visualized in a suitable two-dimensional parameter/state space. A comparative study of the homogeneous equilibrium model (HEM) and the drift flux model (DFM) is carried out to investigate the effects of the DFM parameters the void distribution parameter C 0 and the drift velocity V gi -on the SB, the nature of PAH bifurcation, and on the type of oscillation mode (in-phase or out-of-phase). (author)

  8. Nonlinear full two-fluid study of m=0 sausage instabilities in an axisymmetric Z pinch

    International Nuclear Information System (INIS)

    Loverich, J.; Shumlak, U.

    2006-01-01

    A nonlinear full five-moment two-fluid model is used to study axisymmetric instabilities in a Z pinch. When the electron velocity due to the current J is greater than the ion acoustic speed, high wave-number sausage instabilities develop that initiate shock waves in the ion fluid. This condition corresponds to a pinch radius on the order of a few ion Larmor radii

  9. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    Behrooz, A.

    2008-01-01

    for a large reactor could reach 20 gigabytes) that it is not possible to load into RAM memory of an operating system with 32 bit architecture. A special procedure has been developed within the MATLAB environment to remove this memory limitation, and to invert such large matrices and finally obtain the reactor transfer functions that enable the study of system stability. Various applications of the present frequency-domain code to a typical BWR fuel assembly, a BWR core, and to a chemical reactor showed a good agreement with reference results. (author)

  10. Comparative study of the loss cone-driven instabilities in the low solar corona

    International Nuclear Information System (INIS)

    Sharma, R.R.; Vlahos, L.

    1984-01-01

    A comparative study of the loss cone--driven instabilities in the low solar corona is undertaken. The instailities considered are the electron maser, the whistler, and the electrostatic upper hybrid. We show that the first-harmonic extraordinary mode of the electron cyclotron maser instability is the fastest growing mode for strongly magnetized plasma (ω/sub e//Ω/sub e/ 1.0, no direct electromagnetic radiation is expected since other instabilities, which do not escape directly, saturate the electron cyclotron maser (the whistler or the electrostatic upper hybrid waves). We also show that the second-harmonic electron cyclotron maser emission never grows to an appreciable level. Thus, we suggest that the electron cyclotron maser instability can be the explanation for intense radio bursts only when the first harmonic escapes from the low corona. We propose a possible explanation for the escape of the first harmonic from a flaring loop

  11. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

    1989-01-01

    BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

  12. Experimental studies of instabilities in radially imploding plasmas

    International Nuclear Information System (INIS)

    Oona, H.; Bowers, R.L.; Peterson, D.L.; Findley, C.; Cohrane, J.C.; Ladish, J.S.

    1992-01-01

    In this paper we present experimental data from two experiments that show the initation and growth of the instabilities and compare them to a two dimensional radiational magnetohydrodynamic (RMHD) computation. High-speed cameras (Imacons) were used to observe the implosion as the plasma converges and stagnates on axis. Computer-image enhancement was used to determine the periodicity and amplitude of the initial perturbations. In the two experimental cases, the 5-cm-radius implosion foils had mass of 4.2 and 12.6 milligrams. The data show that the larger mass evolves into a more uniform implosion. Also, we determine the implosion velocity and the growth rate of a selected perturbation

  13. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  14. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  15. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  16. Development of next BWR plant

    International Nuclear Information System (INIS)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

    1995-01-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

  17. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1995-04-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

  18. BWR ATWS mitigation by Fine Motion Control Rod

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

    1994-01-01

    Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

  19. Carpal instability

    International Nuclear Information System (INIS)

    Schmitt, R.; Froehner, S.; Coblenz, G.; Christopoulos, G.

    2006-01-01

    This review addresses the pathoanatomical basics as well as the clinical and radiological presentation of instability patterns of the wrist. Carpal instability mostly follows an injury; however, other diseases, like CPPD arthropathy, can be associated. Instability occurs either if the carpus is unable to sustain physiologic loads (''dyskinetics'') or suffers from abnormal motion of its bones during movement (''dyskinematics''). In the classification of carpal instability, dissociative subcategories (located within proximal carpal row) are differentiated from non-dissociative subcategories (present between the carpal rows) and combined patterns. It is essential to note that the unstable wrist initially does not cause relevant signs in standard radiograms, therefore being ''occult'' for the radiologic assessment. This paper emphasizes the high utility of kinematographic studies, contrast-enhanced magnetic resonance imaging (MRI) and MR arthrography for detecting these predynamic and dynamic instability stages. Later in the natural history of carpal instability, static malalignment of the wrist and osteoarthritis will develop, both being associated with significant morbidity and disability. To prevent individual and socio-economic implications, the handsurgeon or orthopedist, as well as the radiologist, is challenged for early and precise diagnosis. (orig.)

  20. Comparative study of the loss cone-driven instabilities in the low solar corona

    Science.gov (United States)

    Sharma, R. R.; Vlahos, L.

    1984-01-01

    A comparative study of the loss cone-driven instabilities in the low solar corona is undertaken. The instabilities considered are the electron cyclotron maser, the whistler, and the electrostatic upper hybrid. It is shown that the first-harmonic extraordinary mode of the electron cyclotron maser instability is the fastest growing mode for strong magnetized plasma (the ratio of plasma frequency to cyclotron frequency being less than 0.35). For values of the ratio between 0.35 and 1.0, the first-harmonic ordinary mode of the electron cyclotron maser instability dominates the emission. For ratio values greater than 1.0, no direct electromagnetic radiation is expected since other instabilities, which do not escape directly, saturate the electron cyclotron maser (the whistler or the electrostatic upper hybrid waves). It is also shown that the second-harmonic electron cyclotron maser emission never grows to an appreciable level. Thus, it is suggested that the electron cyclotron maser instability can be the explanation for the escape of the first harmonic from a flaring loop.

  1. An analytic study of the perpendicularly propagating electromagnetic drift instabilities in the Magnetic Reconnection Experiment

    International Nuclear Information System (INIS)

    Wang Yansong; Kulsrud, Russell; Ji, Hantao

    2008-01-01

    A local linear theory is proposed for a perpendicularly propagating drift instability driven by relative drifts between electrons and ions. The theory takes into account local cross-field current, pressure gradients, and modest collisions as in the Magnetic Reconnection Experiment [M. Yamada et al., Phys. Plasmas 4, 1936 (1997)]. The unstable waves have very small group velocities in the direction of the pressure gradient, but have a large phase velocity near the relative drift velocity between electrons and ions in the direction of the cross-field current. By taking into account the electron-ion collisions and applying the theory in the Harris sheet, we establish that this instability could be excited near the center of the Harris sheet and have enough e-foldings to grow to large amplitude before it propagates out of the unstable region. Comparing with the other magnetic reconnection related instabilities (lower-hybrid-drift instability, modified two-stream instability, etc.) studied previously, we believe the instability we found is a favorable candidate to produce anomalous resistivity because of its unique wave characteristics, such as electromagnetic component, large phase velocity, and small group velocity in the cross-current-layer direction.

  2. Theoretical and numerical study of Rayleigh-Taylor instabilities in magnetized plasmas

    International Nuclear Information System (INIS)

    Andrei, A. Ivanov

    2001-06-01

    In this thesis we're studying both the general case of the 'classic' Rayleigh-Taylor instability (in incompressible fluids) and more specific cases of the instabilities of Rayleigh-Taylor type in magnetized plasmas, in the liners or wire array implosions etc. We have studied the influence of the Hall diffusion of magnetic field on the growth rate of the instability. We have obtained in this work a self-similar solution for the widening of the initial profile of the magnetic field and for the wave of the penetration of magnetic field. After that the subsequent evolution of the magnetic field in plasma opening switches (POS) has been examined. We have shown the possibility of the existence of a strong rarefaction wave for collisional and non-collisional cases. This wave can explain the phenomenon of the opening of POS. The effect of the suppression of Rayleigh-Taylor instability by forced oscillations of the boundary between two fluids permits us to propose some ideas for the experiments of inertial fusion. We have considered the general case of the instability, in other words - two incompressible viscous superposed fluids in a gravitational field. We have obtained an exact analytical expression for the growth rate and then we have analyzed the influence of the parameters of external 'pumping' on the instability. These results can be applied to a wide range of systems, starting from classic hydrodynamics and up to astrophysical plasmas. The scheme of wire arrays has become recently a very popular method to obtain a high power X-radiation or for a high quality implosion in Z-pinches. The experimental studies have demonstrated that the results of implosion are much better for the case of multiple thin wires situated cylindrically than in a usual liner scheme. We have examined the problem modeling the stabilization of Rayleigh-Taylor instability for a wire array system. The reason for instability suppression is the regular spatial modulation of the surface plasma

  3. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

  4. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

    The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  5. Application of autoregressive methods and Lyapunov coefficients for instability studies of nuclear reactors; Aplicação de métodos autorregressivos e coeficientes de Lyapunov para estudos de instabilidades em reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Aruquipa Coloma, Wilmer

    2017-07-01

    Nuclear reactors are susceptible to instability, causing oscillations in reactor power in specific working regions characterized by determined values of power and coolant mass flow. During reactor startup, there is a greater probability that these regions of instability will be present; another reason may be due to transient processes in some reactor parameters. The analysis of the temporal evolution of the power reveals a stable or unstable process after the disturbance in a light water reactor of type BWR (Boiling Water Reactor). In this work, the instability problem was approached in two ways. The first form is based on the ARMA (Autoregressive Moving Average models) model. This model was used to calculate the Decay Ratio (DR) and natural frequency (NF) of the oscillations, parameters that indicate if the one power signal is stable or not. In this sense, the DRARMA code was developed. In the second form, the problems of instability were analyzed using the classical concepts of non-linear systems, such as Lyapunov exponents, phase space and attractors. The Lyapunov exponents quantify the exponential divergence of the trajectories initially close to the phase space and estimate the amount of chaos in a system; the phase space and the attractors describe the dynamic behavior of the system. The main aim of the instability phenomena studies in nuclear reactors is to try to identify points or regions of operation that can lead to power oscillations conditions. The two approaches were applied to two sets of signals. The first set comes from signals of instability events of the commercial Forsmark reactors 1 and 2 and were used to validate the DRARMA code. The second set was obtained from the simulation of transient events of the Peach Bottom reactor; for the simulation, the PARCS and RELAP5 codes were used for the neutronic/thermal hydraulic coupling calculation. For all analyzes made in this work, the Matlab software was used due to its ease of programming and

  6. 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).

  7. Contribution to study of interfaces instabilities in plane, cylindrical and spherical geometry

    Science.gov (United States)

    Toque, Nathalie

    1996-12-01

    This thesis proposes several experiments of hydrodynamical instabilities which are studied, numerically and theoretically. The experiments are in plane and cylindrical geometry. Their X-ray radiographies show the evolution of an interface between two solid media crossed by a detonation wave. These materials are initially solid. They become liquide under shock wave or stay between two phases, solid and liquid. The numerical study aims at simulating with the codes EAD and Ouranos, the interfaces instabilities which appear in the experiments. The experimental radiographies and the numerical pictures are in quite good agreement. The theoretical study suggests to modelise a spatio-temporal part of the experiments to obtain the quantitative development of perturbations at the interfaces and in the flows. The models are linear and in plane, cylindrical and spherical geometry. They preceed the inoming study of transition between linear and non linear development of instabilities in multifluids flows crossed by shock waves.

  8. Numerical studies of the Kelvin-Hemholtz instability in a coronal jet

    Science.gov (United States)

    Zhao, Tian-Le; Ni, Lei; Lin, Jun; Ziegler, Udo

    2018-04-01

    Kelvin-Hemholtz (K-H) instability in a coronal EUV jet is studied via 2.5D MHD numerical simulations. The jet results from magnetic reconnection due to the interaction of the newly emerging magnetic field and the pre-existing magnetic field in the corona. Our results show that the Alfvén Mach number along the jet is about 5–14 just before the instability occurs, and it is even higher than 14 at some local areas. During the K-H instability process, several vortex-like plasma blobs with high temperature and high density appear along the jet, and magnetic fields have also been rolled up and the magnetic configuration including anti-parallel magnetic fields forms, which leads to magnetic reconnection at many X-points and current sheet fragments inside the vortex-like blob. After magnetic islands appear inside the main current sheet, the total kinetic energy of the reconnection outflows decreases, and cannot support the formation of the vortex-like blob along the jet any longer, then the K-H instability eventually disappears. We also present the results about how the guide field and flux emerging speed affect the K-H instability. We find that a strong guide field inhibits shock formation in the reconnecting upward outflow regions but helps secondary magnetic islands appear earlier in the main current sheet, and then apparently suppresses the K-H instability. As the speed of the emerging magnetic field decreases, the K-H instability appears later, the highest temperature inside the vortex blob gets lower and the vortex structure gets smaller.

  9. Experimental study of the Richtmyer-Meshkov instability induced by a Mach 3 shock wave

    International Nuclear Information System (INIS)

    BP Puranik; JG Oakley; MH Anderson; R Bonaazza

    2003-01-01

    OAK-B135 An experimental investigation of a shock-induced interfacial instability (Richtmyer-Meshkov instability) is undertaken in an effort to study temporal evolution of interfacial perturbations in the late stages of development. The experiments are performed in a vertical shock tube with a square cross-section. A membraneless interface is prepared by retracting a sinusoidally shaped metal plate initially separating carbon dioxide from air, with both gases initially at atmospheric pressure. With carbon dioxide above the plate, the Rayleigh-Taylor instability commences as the plate is retracted and the amplitude of the initial sinusoidal perturbation imposed on the interface begins to grow. The interface is accelerated by a strong shock wave (M=3.08) while its shape is still sinusoidal and before the Kelvin-Helmhotz instability distorts it into the well known mushroom-like structures; its initial amplitude to wavelength ratio is large enough that the interface evolution enters its nonlinear stage very shortly after shock acceleration. The pre-shock evolution of the interface due to the Rayleigh-Taylor instability and the post-shock evolution of the interface due to the Richtmyer-Meshkov instability are visualized using planar Mie scattering. The pre-shock evolution of the interface is carried out in an independent set of experiments. The initial conditions for the Richtmyer-Meshkov experiment are determined from the pre-shock Rayleigh-Taylor growth. One image of the post-shock interface is obtained per experiment and image sequences, showing the post-shock evolution of the interface, are constructed from several experiments. The growth rate of the perturbation amplitude is measured and compared with two recent analytical models of the Richtmyer-Meshkov instability

  10. Theoretical and numerical studies of Rayleigh-Taylor instabilities in magnetized plasmas

    International Nuclear Information System (INIS)

    Ivanov, A.A.

    2001-06-01

    The instabilities of Rayleigh-Taylor type are considered in the thesis. The topic of the thesis was inspired by recent advances in the physics of plasma compression, especially with the aid of systems like Z-pinch. Rayleigh-Taylor instability (RTI) plays an important role in the evolution of magnetized plasmas in these experiments, as well as in stellar plasmas and classic fluids. For the phenomena concerning the nuclear fusion the RTI is very often the factor limiting the possibility of compression. In the current work we try to examine in detail the characteristic features of the instabilities of this type in order to eliminate their detrimental influence. In this thesis we are studying both the general case of the 'classic' Rayleigh-Taylor instability (in incompressible fluids) and more specific cases of the instabilities of Rayleigh-Taylor type in magnetized plasmas, in the liners or wire array implosions etc. We have studied the influence of the Hall diffusion of magnetic field on the growth rate of the instability. We have obtained in this work a self-similar solution for the widening of the initial profile of the magnetic field and for the wave of the penetration of magnetic field. After that the subsequent evolution of the magnetic field in plasma opening switches (POS) has been examined. We have shown the possibility of the existence of a strong rarefaction wave for collisional and non-collisional cases. This wave can explain the phenomenon of the opening of POS. The effect of the suppression of Rayleigh-Taylor instability by forced oscillations of the boundary between two fluids permits us to propose some ideas for the experiments of inertial fusion. We have considered the general case of the instability, in other words, two incompressible viscous superposed fluids in a gravitational field. We have obtained an exact analytical expression for the growth rate and then we have analyzed the influence of the parameters of external 'pumping' on the instability

  11. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  12. An Experimental Study of the Turbulent Development of Rayleigh-Taylor and Richtmyer-Meshkov Instabilities

    International Nuclear Information System (INIS)

    Jacobs, Jeffrey W.

    2006-01-01

    The objective of this three-year research program is to study the development of turbulence in Rayleigh-Taylor (RT) and Richtmyer-Meshkov (RM) instabilities. Incompressible RT and RM instabilities are studied in an apparatus in which a box containing two unequal density liquids is accelerated on a linear rail system either impulsively (by bouncing it off of a spring) to produce RM instability, or at a constant downward rate (using a weight and pulley system) to produce RT instability. These experiments are distinguished from others in the field in that they are initialized with well defined, measurable initial perturbations and are well visualized utilizing planar laser induced fluorescence imaging. New experiments are proposed aimed at generating fully turbulent RM and RT instabilities and quantifying the turbulent development once fully turbulent flows are achieved. The proposed experiments focus on the development and the subsequent application of techniques to accelerate the production of fully turbulent instabilities and the quantification of the turbulent instabilities once they are achieved. The proposed tasks include: the development of RM and RT experiments utilizing fluid combinations having larger density ratios than those previously used; the development of RM experiments with larger acceleration impulse than that previously used; and the investigation of the multi-mode and three-dimensional instabilities by the development of new techniques for generating short wavelength initial perturbations. Progress towards fulfilling these goals is currently well on track. Recent results have been obtained on experiments that utilize Faraday resonance for the production of a nearly single-mode three-dimensional perturbation with a short enough wavelength to yield a self-similar instability at late-times. Last year we reported that we can reliably generate Faraday internal waves on the interface in our experimental apparatus by oscillating the tank containing the

  13. Study of an instability of the PEP-II positron beam (Ohmi effect and Multipactoring)

    Energy Technology Data Exchange (ETDEWEB)

    Heifets, S A [Stanford Linear Accelerator Center, Menlo Park, CA (United States)

    1996-08-01

    The processes defining the density distribution of the photoelectrons are quite complicated. In this study, a simplified model of the instability was used to get a quick estimate of the growth rate of the instability and the relative importance of the parameters, as has been done in Ohmi`s paper. The production rate and dynamics of the photoelectrons are studied for the PEP-II parameters. The growth rate of the transverse instability driven by the primary photoelectrons is of the order of 0.7 msec for the PEP-II parameters. The multipactoring at resonance currents cannot produce large electron density due to the final energy spread caused by the finite bunch length and the intrinsic energy spread of the secondary electrons. Production of the secondary electrons may lead to large average densities. The ion can be produced in electron collisions with the residual gas with density of the order of the electron density. (G.K.)

  14. Computational and experimental studies of hydrodynamic instabilities and turbulent mixing (Review of NVIIEF efforts)

    International Nuclear Information System (INIS)

    Andronov, V.A.; Zhidov, I.G.; Meskov, E.E.; Nevmerzhitskii, N.V.; Nikiforov, V.V.; Razin, A.N.; Rogatchev, V.G.; Tolshmyakov, A.I.; Yanilkin, Yu.V.

    1995-02-01

    This report describes an extensive program of investigations conducted at Arzamas-16 in Russia over the past several decades. The focus of the work is on material interface instability and the mixing of two materials. Part 1 of the report discusses analytical and computational studies of hydrodynamic instabilities and turbulent mixing. The EGAK codes are described and results are illustrated for several types of unstable flow. Semiempirical turbulence transport equations are derived for the mixing of two materials, and their capabilities are illustrated for several examples. Part 2 discusses the experimental studies that have been performed to investigate instabilities and turbulent mixing. Shock-tube and jelly techniques are described in considerable detail. Results are presented for many circumstances and configurations

  15. A numerical study of boiling flow instability of a reactor thermosyphon system

    International Nuclear Information System (INIS)

    Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der; Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew

    2006-01-01

    A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed

  16. A numerical study of boiling flow instability of a reactor thermosyphon system

    Energy Technology Data Exchange (ETDEWEB)

    Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der [Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew [Shell Research and Technology Centre, Badhuisweg 3, 1031 CM Amsterdam (Netherlands)

    2006-04-01

    A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed. [Author].

  17. Longitudinal instability studies at the SURF II storage ring at NIST

    International Nuclear Information System (INIS)

    Harkay, K.C.; Sereno, N.S.

    1998-01-01

    Measurements of the longitudinal instability observed in the storage ring at the Synchrotrons Ultraviolet Radiation Facility (SURF II) at the National Institute of Standards and Technology (NET) were performed to understand the mechanism driving the instability. The instability, studied in depth by Ralcowsky and others, manifests itself in broad resonance features in the horizontal and vertical motion spectrum of the synchrotrons light from DC to a few kHz. Also observed are multiple synchrotrons harmonics that modulate the revolution harmonics; these are characteristic of longitudinal phase oscillations. These spectral features of the motion are found to be correlated with the periodic lengthening and shortening of the bunch length on time scales from approximately0.1 ms to 20 ms, depending on machine and radio-frequency (rf) system parameters. In this report, the growth rate of the instability is determined from measurements using an rf pickup electrode. The measured growth rates are compared to computed growth rates from an analytical model. Recommendations are made regarding options to control or mitigate the instability. In light of upgrade plans for SURF III, a few comments are presented about the beam lifetime

  18. A study on the effect of macroeconomics instability index on private investment in Iran

    Directory of Open Access Journals (Sweden)

    Aziz Saki

    2012-10-01

    Full Text Available In this paper, we perform an empirical study to investigate the impact of economical stability on the amount of investment coming from the private sector. We calculate macroeconomics instability index (MIX using the existing methods in the literature. We have also used Glezakos (1973 method [Glezakos,C.(1973. Export instability and economic growth: A statistical verification. Economic Development and Cultural Change, 21(3, 670-678.], which considers long-term deviation of real values as instability index. Therefore, we use four variables of inflation rate (TINF, the ratio of budget deficit on growth domestic product (GDP (TBD, foreign debt on GDP (TFD and the ratio of actual currency rate on nominate currency (TFD. The preliminary results show that the short-term changes on LNIP with one lag and LNIV have positive impact on LNIP. In addition, any short term changes on LNMII has negative and meaningful impact on LNIP and approximately 0.67 percent of difference between the actual and long term are discounted in each period. The results indicate that instability index has negative effect even in short term on Iran's industry. This shows the relevant importance of instability on economy.

  19. Evaluation of internal flooding in a BWR

    International Nuclear Information System (INIS)

    Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

    1985-01-01

    Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event

  20. BWR Mark I pressure suppression study: characterization of the vertical load function utilizing bench top model tests

    International Nuclear Information System (INIS)

    McCauley, E.W.; Lai, W.

    1977-02-01

    A study was conducted to characterize the mechanisms which give rise to observed oscillations in the vertical load function (VLF) of bench top pool dynamics tests. This is part of a continuing investigation at the Lawrence Livermore Laboratory of the General Electric Mark I Nuclear Reactor pressure suppression system

  1. Patellar Instability Management: A Survey of the International Patellofemoral Study Group.

    Science.gov (United States)

    Liu, Joseph N; Steinhaus, Michael E; Kalbian, Irene L; Post, William R; Green, Daniel W; Strickland, Sabrina M; Shubin Stein, Beth E

    2017-10-01

    Although patellofemoral instability is among the most prevalent knee disorders, the management of patients with this condition is complex and remains variable, given the lack of long-term, high-level clinical outcome studies to compare various operative and nonoperative modalities. To discover a consensus within treatment controversies in patellofemoral instability among experienced knee surgeons with a specific interest in the patellofemoral joint. Expert opinion; Level of evidence, 5. A 3-step modified Delphi technique was used to establish a consensus. A 34-question, case-based online survey regarding patellofemoral instability was distributed to all active members of the International Patellofemoral Study Group. Consensus statements were generated if at least 66% of the respondents agreed and then redistributed to the same panel. Modifications to the consensus statements were made based on the iterative feedback process until no discordance was encountered in the third stage. Eight consensus statements were achieved. Nonoperative management is the current standard of care for a first-time dislocation in the absence of an osteochondral fragment or loose body requiring excision (100% agreement). In patients with a first-time dislocation with an operative osteochondral fracture requiring excision or repair, patellar instability should be addressed concurrently (89% agreement). Recurrent instability should be treated surgically, with most surgeons favoring medial reconstruction (77%-86% agreement). While there is general agreement that bony procedures should be performed to correct underlying bony deformities, there is no consensus regarding the most appropriate type of procedure performed. Lateral release should not be performed in isolation for the treatment of patellar instability (89% agreement). Despite the consensus generated in this study, our current understanding remains limited by a lack of high-level evidence as well as the numerous complex variables

  2. Study on the dimensional instability of metallic uranium subject to thermal alternation

    International Nuclear Information System (INIS)

    Gentile, E.F.

    1976-01-01

    Methalographic properties of metallic uranium submitted to a thermal cycle are studied. Microstructures heat treatment and methods utilized are presented. Dimensional instability of uranium is the main subject of the study and it is seen that it is strongly reduced in the presence of molybdenum [pt

  3. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

  4. Study on MHD instabilities in the CECI plasma device using Fourier probes

    International Nuclear Information System (INIS)

    Rosal, A.C.; Aso, Y.; Ueda, M.

    1991-01-01

    A magnetic diagnostics called Fourier analyser aiming to study MHD instabilities by Fourier series expansion of poloidal magnetic field for m ≤ 3 modes was developed and tested. The diagnostics will be used in the RFP (reversed field pinch) type toroidal plasma device. (M.C.K.)

  5. BWR steel containment corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Tan, C.P.; Bagchi, G.

    1996-04-01

    The report describes regulatory actions taken after corrosion was discovered in the drywell at the Oyster Creek Plant and in the torus at the Nine Mile Point 1 Plant. The report describes the causes of corrosion, requirements for monitoring corrosion, and measures to mitigate the corrosive environment for the two plants. The report describes the issuances of generic letters and information notices either to collect information to determine whether the problem is generic or to alert the licensees of similar plants about the existence of such a problem. Implementation of measures to enhance the containment performance under severe accident conditions is discussed. A study by Brookhaven National Laboratory (BNL) of the performance of a degraded containment under severe accident conditions is summarized. The details of the BNL study are in the appendix to the report.

  6. BWR alloy 182 stress Corrosion Cracking Experience

    International Nuclear Information System (INIS)

    Horn, R.M.; Hickling, J.

    2002-01-01

    Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

  7. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  8. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  9. Magnetohydrodynamic study for three-dimensional instability of the Petschek type magnetic reconnection

    International Nuclear Information System (INIS)

    Shimizu, T.; Kondoh, K.

    2013-01-01

    The 3D instability of the spontaneous fast magnetic reconnection process is studied with magnetohydrodynamics (MHD) simulations, where the 2D model of the spontaneous fast magnetic reconnection is destabilized in three dimension. As well known in many 2D numerical MHD studies, when a 1D current sheet is destabilized with the current-driven anomalous resistivity, the 2D Petschek type fast magnetic reconnection is established. This paper shows that the 2D Petschek type fast magnetic reconnection can be destabilized in three dimension by an initial resistive disturbance which includes a weak fluctuation in the sheet current direction, i.e., along the magnetic neutral line. The resulting 3D fast magnetic reconnection finally becomes intermittent and random through a 3D instability. In addition, it is also shown that the 3D instability is suppressed by the uniform resistivity. It suggests that the 3D instability is caused in the Petschek-type reconnection process which is characterized by a strongly localized magnetic diffusion region and the slow shock acceleration of the plasma jets and is suppressed in the Sweet-Parker type reconnection process

  10. Linear and nonlinear studies of velocity shear driven three dimensional electron-magnetohydrodynamics instability

    International Nuclear Information System (INIS)

    Gaur, Gurudatt; Das, Amita

    2012-01-01

    The study of electron velocity shear driven instability in electron magnetohydrodynamics (EMHD) regime in three dimensions has been carried out. It is well known that the instability is non-local in the plane defined by the flow direction and that of the shear, which is the usual Kelvin-Helmholtz mode, often termed as the sausage mode in the context of EMHD. On the other hand, a local instability with perturbations in the plane defined by the shear and the magnetic field direction exists which is termed as kink mode. The interplay of these two modes for simple sheared flow case as well as that when an external magnetic field exists has been studied extensively in the present manuscript in both linear and nonlinear regimes. Finally, these instability processes have been investigated for the exact 2D dipole solutions of EMHD equations [M. B. Isichenko and A. N. Marnachev, Sov. Phys. JETP 66, 702 (1987)] for which the electron flow velocity is sheared. It has been shown that dipoles are very robust and stable against the sausage mode as the unstable wavelengths are typically longer than the dipole size. However, we observe that they do get destabilized by the local kink mode.

  11. Parametric study on kink instabilities of twisted magnetic flux ropes in the solar atmosphere

    Science.gov (United States)

    Mei, Z. X.; Keppens, R.; Roussev, I. I.; Lin, J.

    2018-01-01

    Aims: Twisted magnetic flux ropes (MFRs) in the solar atmosphere have been researched extensively because of their close connection to many solar eruptive phenomena, such as flares, filaments, and coronal mass ejections (CMEs). In this work, we performed a set of 3D isothermal magnetohydrodynamic (MHD) numerical simulations, which use analytical twisted MFR models and study dynamical processes parametrically inside and around current-carrying twisted loops. We aim to generalize earlier findings by applying finite plasma β conditions. Methods: Inside the MFR, approximate internal equilibrium is obtained by pressure from gas and toroidal magnetic fields to maintain balance with the poloidal magnetic field. We selected parameter values to isolate best either internal or external kink instability before studying complex evolutions with mixed characteristics. We studied kink instabilities and magnetic reconnection in MFRs with low and high twists. Results: The curvature of MFRs is responsible for a tire tube force due to its internal plasma pressure, which tends to expand the MFR. The curvature effect of toroidal field inside the MFR leads to a downward movement toward the photosphere. We obtain an approximate internal equilibrium using the opposing characteristics of these two forces. A typical external kink instability totally dominates the evolution of MFR with infinite twist turns. Because of line-tied conditions and the curvature, the central MFR region loses its external equilibrium and erupts outward. We emphasize the possible role of two different kink instabilities during the MFR evolution: internal and external kink. The external kink is due to the violation of the Kruskal-Shafranov condition, while the internal kink requires a safety factor q = 1 surface inside the MFR. We show that in mixed scenarios, where both instabilities compete, complex evolutions occur owing to reconnections around and within the MFR. The S-shaped structures in current distributions

  12. An Analytic Study of the Perpendicularly Propagating Electromagnetic Drift Instabilities in the Magnetic Reconnection Experiment

    International Nuclear Information System (INIS)

    Wang, Y.; Kulsrud, R.; Ji, H.

    2008-01-01

    A local linear theory is proposed for a perpendicularly propagating drift instability driven by relative drifts between electrons and ions. The theory takes into account local cross-field current, pressure gradients and modest collisions as in the Magnetic Reconnection Experiment (MRX) (10). The unstable waves have very small group velocities in the direction of the pressure gradient, but have a large phase velocity near the relative drift velocity between electrons and ions in the direction of cross-field current. By taking into account the electron-ion collisions and applying the theory in the Harris sheet, we establish that this instability could be excited near the center of the Harris sheet and have enough efoldings to grow to large amplitude before it propagates out of the unstable region. Comparing with the other magnetic reconnection related instabilities (LHDI, MTSI et.) studied previously, we believe the instability we find is a favorable candidate to produce anomalous resistivity because of its unique wave characteristics, such as electromagnetic component, large phase velocity, and small group velocity in the cross current layer direction

  13. An Experimental Study of Roughness-Induced Instabilities in a Supersonic Boundary Layer

    Science.gov (United States)

    Kegerise, Michael A.; King, Rudolph A.; Choudhari, Meelan; Li, Fei; Norris, Andrew

    2014-01-01

    Progress on an experimental study of laminar-to-turbulent transition induced by an isolated roughness element in a supersonic laminar boundary layer is reported in this paper. Here, the primary focus is on the effects of roughness planform shape on the instability and transition characteristics. Four different roughness planform shapes were considered (a diamond, a circle, a right triangle, and a 45 degree fence) and the height and width of each one was held fixed so that a consistent frontal area was presented to the oncoming boundary layer. The nominal roughness Reynolds number was 462 and the ratio of the roughness height to the boundary layer thickness was 0.48. Detailed flow- field surveys in the wake of each geometry were performed via hot-wire anemometry. High- and low-speed streaks were observed in the wake of each roughness geometry, and the modified mean flow associated with these streak structures was found to support a single dominant convective instability mode. For the symmetric planform shapes - the diamond and circular planforms - the instability characteristics (mode shapes, growth rates, and frequencies) were found to be similar. For the asymmetric planform shapes - the right-triangle and 45 degree fence planforms - the mode shapes were asymmetrically distributed about the roughness-wake centerline. The instability growth rates for the asymmetric planforms were lower than those for the symmetric planforms and therefore, transition onset was delayed relative to the symmetric planforms.

  14. 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).

  15. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  16. Study of plasma-maser instability in an inhomogeneous plasma

    International Nuclear Information System (INIS)

    Singh, Mahinder

    2006-01-01

    The plasma-maser, an interesting nonlinear process in plasmas, is an effective means of energy up-conversion in frequency from low-frequency turbulence to a high-frequency wave. A theoretical study is made of the amplification mechanism of an electrostatic Bernstein mode wave in presence of Langmuir wave turbulence in a magnetized inhomogeneous plasma on the basis of a plasma-maser interaction. It is shown that a test high-frequency electrostatic Bernstein mode wave is unstable in the presence of low-frequency Langmuir wave turbulence. The growth rate of a test high-frequency Bernstein mode wave is calculated with the involvement of a spatial density gradient parameter. A comparative study on the role of density gradient in the generation of the Bernstein mode on the basis of the plasma-maser effect is presented

  17. Instabilities in inhomogeneous plasma

    International Nuclear Information System (INIS)

    Mikhailovsky, A.B.

    1983-01-01

    The plasma inhomogeneity across the magnetic field causes a wide class of instabilities which are called instabilities of an inhomogeneous plasma or gradient instabilities. The instabilities that can be studied in the approximation of a magnetic field with parallel straight field lines are treated first, followed by a discussion of the influence of shear on these instabilities. The instabilities of a weakly inhomogeneous plasma with the Maxwellian velocity distribution of particles caused by the density and temperature gradients are often called drift instabilities, and the corresponding types of perturbations are the drift waves. An elementary theory of drift instabilities is presented, based on the simplest equations of motion of particles in the field of low-frequency and long-wavelength perturbations. Following that is a more complete theory of inhomogeneous collisionless plasma instabilities which uses the permittivity tensor and, in the case of electrostatic perturbations, the scalar of permittivity. The results are used to study the instabilities of a strongly inhomogeneous plasma. The instabilities of a plasma in crossed fields are discussed and the electromagnetic instabilities of plasma with finite and high pressure are described. (Auth.)

  18. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    International Nuclear Information System (INIS)

    Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

    2008-01-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  19. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

    2008-07-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  20. Experimental study of an ion cyclotron instability in a magnetic well confined plasma

    International Nuclear Information System (INIS)

    Brossier, P.

    1969-01-01

    This report is a contribution to the study of microinstabilities in macroscopically stable plasmas, in the low-β limit. Ion cyclotron instabilities, with k || = 0, have been numerically studied in detail; the computation of the density thresholds and growth rates of the different harmonics showed the relative role played by the following energy sources: density gradient, perpendicular distribution function and cold plasma component. This theoretical model has been compared with the results of a detailed study (density thresholds, wave structure, frequency spectrum, wavelengths, growth rate, amplitude of the electric field) of the instability observed in the DECA II device. This comparison gave a good agreement which shows the destabilising role played by the cold plasma component on a hot plasma with a loss cone distribution function. (author) [fr

  1. Shock wave, fluid instability and implosion studies with a kinetic particle approach

    Science.gov (United States)

    Sagert, Irina; Even, Wesley P.; Strother, Terrance T.

    2016-10-01

    Many problems in laboratory plasma physics are subject to flows that move between the continuum and the kinetic regime. The correct description of these flows is crucial in order to capture their impact on the system's dynamical evolution. Examples are capsule implosions in inertial confinement fusion (ICF). Although their dynamics is predominantly shaped by shock waves and fluid instabilities, non-equilibrium flows in form of deuterium/tritium ions have been shown to play a significant role. We present recent studies with our Monte Carlo kinetic particle code that is designed to capture continuum and kinetic flows in large physical systems with possible applications in ICF studies. Discussed results will include standard shock wave and fluid instability tests and simulations that are adapted towards future ICF studies with comparisons to hydrodynamic simulations. This work used the Wolf TriLAB Capacity Cluster at LANL. I.S. acknowledges support through a Director's fellowship (20150741PRD3) from Los Alamos National Laboratory.

  2. 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)

  3. Study of the Rayleigh-Taylor instability at the ablation front

    International Nuclear Information System (INIS)

    Salvatore, Patricia

    2000-01-01

    Inertial confinement fusion in indirect drive consists in irradiating with ultra powerful laser beams the internal wall of a heating cavity which contains a capsule enclosing the thermonuclear fuel. During laser-matter interaction, laser light is converted into x-rays onto the hohlraum walls. The x-rays capsule heating produces a matter expansion, this one induces a pressure accelerating the capsule wall which implodes and compresses the fuel. The limit between the expanded plasma and the accelerated one is named ablation front. A light fluid (the ablated plasma) accelerating a heavy one (the shell) seeds Rayleigh-Taylor instability. To perform experiments, we used the Phebus facility at Limeil-Valenton CEA (the most powerful laser in Europe). After frequency conversion, each laser beam can deliver onto a target an energy up to 3 kJ at 0.35μm wavelength. In the United States of America and in France, more powerful laser facilities are planned to deliver an energy about 1 MJ: the National Ignition Facility (Lawrence Livermore National Laboratory, California) and the Laser MegaJoule (CEA, Bordeaux). Hydrodynamic instabilities take an important part in the definition of these facilities. Two main experiments were carried out on the Phebus laser. We studied the Rayleigh-Taylor instability at the ablation front with a modulated CHBr plane target stuck on the gold hohlraum wall. During the september-october 1996 experiment, a x-ray device was used. We observed the temporal evolution of the target modulations by x-ray imaging cinematography which recorded face-on radiographs. The second experiment was performed with collaboration of the Imperial College of London. Two high spatial resolution devices (less than 5 μm) were used in order to study short wavelengths modulations. The first diagnostic recorded side-on observations of target acceleration, the second one was used to measure the instability growth with face-on radiography. We studied this growth in a modulation

  4. Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments; TOPICAL

    International Nuclear Information System (INIS)

    Ott, L.J.

    1994-01-01

    The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments[Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a basis

  5. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

    2006-01-01

    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 Laguna Verde

  6. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake

    1995-01-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

  7. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

    1995-03-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

  8. A urodynamic study of surface neuromodulation versus sham in detrusor instability and sensory urgency.

    Science.gov (United States)

    Bower, W F; Moore, K H; Adams, R D; Shepherd, R

    1998-12-01

    We studied the effect of surface neuromodulation on cystometric pressure and volume parameters in women with detrusor instability or sensory urgency. Electrical current was delivered to the suprapubic region and third sacral foramina via a transcutaneous electrical nerve stimulator with sham neuromodulation control. A consecutive series of women with proved detrusor instability or sensory urgency were randomized to 3 surface neuromodulation groups. Volume and pressure parameters were the main outcomes of transcutaneous electrical nerve stimulation applied during second cystometric fill. Sham transcutaneous electrical nerve stimulation did not alter the outcome measures. However, neuromodulation delivered across the suprapubic and sacral skin effected a reduction in mean maximum height of detrusor contraction. A current which inhibits motor activity was not superior to that which inhibits sensory perception in reducing detrusor pressure. Response in sensory urgency was poor. Results from our sham controlled study suggest that short-term surface neuromodulation via transcutaneous electrical nerve stimulation may have a role in the treatment of detrusor instability. Future studies must examine the clinical effect of long-term surface neuromodulation.

  9. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  10. Study of parametric instabilities during the Alcator C lower hybrid wave heating experiments

    International Nuclear Information System (INIS)

    Takase, Y.

    1983-10-01

    Parametric excitation of ion-cyclotron quasi-modes (ω/sub R/ approx. = nω/sub ci/) and ion-sound quasi-modes (ω/sub R/ approx. = k/sub parallel to/v/sub ti/) during lower hybrid wave heating of tokamak plasmas have been studied in detail. Such instabilities may significantly modify the incident wavenumber spectrum near the plasma edge. Convective losses for these instabilities are high if well-defined resonance cones exist, but they are significantly reduced if the resonance cones spread and fill the plasma volume (or some region of it). These instabilities preferentially excite lower hybrid waves with larger values of n/sub parallel to/ than themselves possess, and the new waves tend to be absorbed near the outer layers of the plasma. Parametric instabilities during lower hybrid heating of Alcator C plasmas have been investigated using rf probes (to study tilde phi and tilde n/sub i/) and CO 2 scattering technique (to study tilde n/sub e/). At lower densities (anti n/sub e/ less than or equal to 0.5 x 10 14 cm -3 ) where waves observed in the plasma interior using CO 2 scattering appear to be localized, parametric decay is very weak. Both ion-sound and ion-cyclotron parametric decay processes have been observed at higher densities (anti n greater than or equal to 1.5 x 10 14 cm -3 ) where waves appear to be unlocalized. Finally, at still higher densities (anti n /sub e/ greater than or equal to 2 x 10 4 cm -3 ) pump depletion has been observed. Above these densities heating and current drive efficiencies are expected to degrade significantly

  11. A Computational Study of Richtmyer-Meshkov Instability with Surface Tension

    Science.gov (United States)

    Francois, Marianne; Velechovsky, Jan; Jibben, Zach; Masser, Thomas; LANL Collaboration

    2017-11-01

    We have added the capability to model surface tension in our adaptive mesh refinement compressible flow solver, xRage. Our surface tension capability employs the continuum surface force to model surface tension and the height function method to compute curvatures. We have verified our model implementation for the static and oscillating droplets test cases and the linear regime of the Rayleigh-Taylor instability. With this newly added capability, we have performed a numerical study of the effects of surface tension on single-mode and multi-mode Richtmyer-Meshkov instability. This work was performed under the auspices of the National Nuclear Security Administration of the U.S. Department of Energy at Los Alamos National Laboratory under Contract No. DE-AC52 - 06NA25396.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  13. A numerical study of three-dimensional bubble merger in the Rayleigh endash Taylor instability

    International Nuclear Information System (INIS)

    Li, X.L.

    1996-01-01

    The Rayleigh endash Taylor instability arises when a heavy fluid adjacent to a light fluid is accelerated in a direction against the density gradient. Under this unstable configuration, a perturbation mode of small amplitude grows into bubbles of the light fluid and spikes of the heavy fluid. Taylor discovered the steady state motion with constant velocity for a single bubble or periodic bubbles in the Rayleigh endash Taylor instability. Read and Youngs studied the motion of a randomly perturbed fluid interface in the Rayleigh endash Taylor instability. They reported constant acceleration for the overall bubble envelope. Bubble merger is believed to cause the transition from constant velocity to constant acceleration. In this paper, we present a numerical study of this important physical phenomenon. It analyzes the physical process of bubble merger and the relationship between the horizontal bubble expansion and the vertical interface acceleration. A dynamic bubble velocity, beyond Taylor close-quote s steady state value, is observed during the merger process. It is believed that this velocity is due to the superposition of the bubble velocity with a secondary subharmonic unstable mode. The numerical results are compared with experiments. copyright 1996 American Institute of Physics

  14. The risk of PCI damage to 8x8 fuel rods during limit cycle instability

    Energy Technology Data Exchange (ETDEWEB)

    Schrire, D.; Oguma, R.; Malen, K.

    1994-12-31

    A BWR reactor core may experience thermal-hydraulic instability under certain operating conditions. Generally, the instability results in neutron flux (i e generated neutronic power) and coolant flow and pressure oscillations, which reach a maximum `limit cycle` amplitude. The cladding response to power transients has been studied using noise analysis. These results have been compared to results from code calculations using the fuel code TOODEE 2. From these results the risk for fuel rod failure due to pellet-clad mechanical interaction and possible failure due to stress corrosion cracking (PCI) has been estimated. It turns out that for the oscillation frequencies of interest (0,3-0,5 Hz) the fuel response amplitude reduction makes PCI-failure improbable. 17 refs.

  15. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  16. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry

    International Nuclear Information System (INIS)

    Moranchel, M.; Garcia B, A.; Longoria G, L. C.

    2012-01-01

    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 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 investigation of

  17. Studies of the trapped particle and ion temperature gradient instabilities in the Columbia Linear Machine

    International Nuclear Information System (INIS)

    Mathey, O.H.

    1989-01-01

    In the first part of the work, the effects of weak Coulomb and neutral collisions on the collisionless curvature driven trapped particle mode are studied in the Columbia Linear Machine (CLM) [Phys. Rev. Lett. 57, 1729, (1986)]. Low Coulomb collisionality yields a small stabilizing correction to the magnetohydrodynamic (MHD) collisionless mode, which scales as v, using the Krook model, and ν ec 1/2 using a Lorentz pitch angle operator. In higher collisionality regimes, both models tend to yield similar scalings. In view of relative high neutral collisionality in CLM, both types of collisionality are then combined, modeling neutral collisions with the conserving Krook and Coulomb collisions with a Lorentz model. The dispersion relation is then integrated over velocity space. This combination yields results in very good accord with the available experimental data. The Ion Temperature Gradient Instability is then investigated. It is shown that anisotropy in gradient has a substantial effect on the ion temperature gradient driven mode. A gradient in the parallel temperature is needed for an instability to occur, and a gradient in the perpendicular temperature gradient further enhances the instability indirectly as long as the frequency of the mode is near ion resonance. The physical reason for this important role difference is presented. The Columbia Linear Machine is being redesigned to produce and identify the ion temperature gradient driven η i mode. Using the expected parameters, the author has developed detailed predictions of the mode characteristics in the CLM. Strong multi mode instabilities are expected. As the ion parallel and perpendicular ion temperature gradients are expected to differ significantly, we differentiate between η i parallel and ν i perpendicular and explore the physical differences between them, which leads to a scheme for stabilization of the mode

  18. Prevalence of chronic ankle instability and associated symptoms in university dance majors: an exploratory study.

    Science.gov (United States)

    Simon, Janet; Hall, Emily; Docherty, Carrie

    2014-01-01

    Previous investigations have established that dancers suffer a large number of injuries to the lower leg, foot, and ankle, with a portion of these being significant time loss injuries or in some cases career ending. Lateral ankle sprain is a common injury in dancers and can often lead to recurrent instability and repetitive injuries. Research in other active populations has linked ankle sprains to the development of chronic ankle instability (CAI). Therefore, the purpose of this study was to identify the prevalence of CAI and related symptoms of ankle sprain in a student dance population. Individuals were included if they were currently a modern or ballet dance major at the investigators' university (exclusion criterion: a history of fracture or surgery in the lower extremities). A self-reported demographic questionnaire and the Identification of Functional Ankle Instability survey were used to identify the presence and characteristics of CAI. A total of 83 questionnaires were collected, and after exclusions, 77 participants remained: 43 modern dancers and 34 ballet dancers (10 males and 67 females, mean age 19.61 ± 2.53 years, mean dance experience 13.61 ± 3.16 years). Of all dancers surveyed, 41 (53.2%) had CAI, and of those 24 (58.5%) were modern dancers, and 17 (41.5%) were ballet dancers. When looking only at those dancers who had a previous lateral ankle sprain, 75.9% were identified as having CAI. Chronic Ankle Instability can create long-term problems for anyone but especially female dancers, who place extreme stress on their feet and ankles from being en pointe or demi-pointe. It is important to educate dancers, instructors, and medical staff of the importance of recognizing CAI and seeking medical care for ankle sprains and their residual symptoms.

  19. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  20. NATO Advanced Study Institute on Instabilities and Chaos in Quantum Optics

    CERN Document Server

    Arecchi, F; Lugiato, L; Instabilities and Chaos in Quantum Optics II

    1988-01-01

    This volume contains tutorial papers from the lectures and seminars presented at the NATO Advanced Study Institute on "Instabilities and Chaos in Quantum Optics", held at the "Il Ciocco" Conference Center, Castelvecchio Pascoli, Lucca, Italy, June 28-July 7, 1987. The title of the volume is designated Instabilities and Chaos in Quantum Optics II, because of the nearly coincident publication of a collection of articles on research in this field edited by F.T. Arecchi and R.G. Harrison [Instabilities and Chaos in Quantum Optics, (Springer, Berlin, 1987) 1. That volume provides more detailed information about some of these topics. Together they will serve as a comprehensive and tutorial pair of companion volumes. This school was directed by Prof. Massimo Inguscio, of the Department of Physics, University of Naples, Naples, Italy to whom we express our gratitude on behalf of all lecturers and students. The Scientific Advisory Committee consisted of N.B. Abraham of Bryn Mawr College; F.T. Arecchi of the National I...

  1. Experimental Studies of the Electrothermal and Magneto-Rayleigh Taylor Instabilities on Thin Metal Foil Ablations

    Science.gov (United States)

    Steiner, Adam; Yager-Elorriaga, David; Patel, Sonal; Jordan, Nicholas; Gilgenbach, Ronald; Lau, Y. Y.

    2015-11-01

    The electrothermal instability (ETI) and magneto-Rayleigh Taylor instability (MRT) are important in the implosion of metallic liners, such as magnetized liner implosion fusion (MagLIF). The MAIZE linear transformer driver (LTD) at the University of Michigan generates 200 ns risetime-current pulses of 500 to 600 kA into Al foil liners to study plasma instabilities and implosion dynamics, most recently MRT growth on imploding cylindrical liners. A full circuit model of MAIZE, along with I-V measurements, yields time-resolved load inductance. This has enabled measurements of an effective current-carrying radius to determine implosion velocity and plasma-vacuum interface acceleration. Measurements are also compared to implosion data from 4-time-frame laser shadowgraphy. Improved resolution measurements on the laser shadowgraph system have been used to examine the liner interface early in the shot to examine surface perturbations resulting from ETI for various seeding conditions. Fourier analysis examines the growth rates of wavelength bands of these structures to examine the transition from ETI to MRT. This work was supported by the U.S. DoE through award DE-SC0012328. S.G. Patel is supported by Sandia National Labs. D.A. Yager is supported by NSF fellowship grant DGE 1256260.

  2. Study of flow instability in a centrifugal fan based on energy gradient theory

    International Nuclear Information System (INIS)

    Xiao, Meina; Dou, Hua-Shu; Ma, Xiaoyang; Xiao, Qing; Chen, Yongning; He, Haijiang; Ye, Xinxue

    2016-01-01

    Flow instability in a centrifugal fan was studied using energy gradient theory. Numerical simulation was performed for the three dimensional turbulent flow field in a centrifugal fan. The flow is governed by the three-dimensional incompressible Navier-Stokes equations coupled with the RNG k-ε turbulent model. The finite volume method was used to discretize the governing equations and the Semiimplicit method for pressure linked equation (SIMPLE) algorithm is employed to iterate the system of the equations. The interior flow field in the centrifugal fan and the distribution of the energy gradient function K are obtained at different flow rates. According to the energy gradient method, the area with larger value of K is the place where the flow loses stability easier. The results show that instability is easier to generate in the regions of impeller outlet and volute tongue. The air flow near the hub is more stable than that near the shroud. That is due to the influences of variations of the velocity and the inlet angle along the axial direction. With the decrease of the flow rate, instability zone in a blade channel moves to the impeller inlet from the outlet and the unstable regions in different channels develop in opposite direction to the rotation of impeller

  3. 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

  4. BWR Refill-Reflood Program. Final report

    International Nuclear Information System (INIS)

    Myers, L.L.

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

  5. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1983-01-01

    A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  6. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  7. Experimental study on dynamic stabilization of the MHD instability in pinch plasmas surrounded by a conducting shell

    International Nuclear Information System (INIS)

    Yamamoto, Shunji; Ishii, Shozo; Kawamoto, Shigeshi; Hayashi, Izumi

    1981-01-01

    Experimental study on the dynamic stabilization of MHD instability with a pinch plasma generator was done, and the results were compared with the theoretical works. The previous results of theoretical analysis showed that a conducting shell worked effectively for the dynamic stabilization of MHD instability. The present experiment was carried out with a linear plasma generator which consisted of a discharge tube, a coil and a conducting shell. The macroscopic behavior of plasma was observed with an image converter camera, and the phenomena due to the instability was measured by a magnetic probe. A sine-cosine coil was employed for the observation of the growth of instability. The following results were obtained. When the frequency of RF current for dynamic stabilization was larger than the growth rate of instability, the experimental results were in agreement with the theoretical ones. The effect of a conducting shell was clearly seen. For the helical instability of short wave length, the dynamic stabilization was easily obtained even without a conducting shell. The self-reversal phenomena due to the helical instability of short wave length was suppressed by the RF current along the axis of a discharge tube. (Kato, T.)

  8. Preliminary study of Rayleigh-Taylor instability in wire-array Z-pinch

    International Nuclear Information System (INIS)

    He Kaihui; Feng Kaiming; Li Qiang; Gao Chunming

    2000-01-01

    It is important to research into the MHD Rayleigh-Taylor instability developed in Z-pinch implosion. A snowplough model of the single wire Z-pinch is presented. The perturbation amplitude of Rayleigh-Taylor instability in the wire-array Z-pinch is analyzed quantitatively. Sheared axial flow is put forward to mitigate and reduce the Rayleigh-Taylor instability. And other approaches used to mitigate MHD instability in such a super-fast process are explored

  9. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  10. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  11. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

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

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  12. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  13. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  14. Experience on a BWR plant diagnosis system

    International Nuclear Information System (INIS)

    Tanabe, A.; Kawai, K.; Hashimoto, Y.

    1981-01-01

    It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

  15. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  16. Analytical study of flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions

    International Nuclear Information System (INIS)

    Nayak, A.K.; Kumar, N.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2002-01-01

    Analytical investigations have been carried out to study the flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions. For this purpose, the computer code TINFLO-S has been developed. The code solves the conservation equations of mass, momentum and energy and equation of state for homogeneous equilibrium twophase flow using linear analytical technique. The results of the code have been validated with the experimental data of the loop for both the steady state and stability. The study reveals that the stability behaviour of low quality flow oscillations is different from that of the high quality flow oscillations. The instability reduces with increase in power and throttling at the inlet of the heater. The instability first increases and then reduces with increase in pressure at any subcooling. The effects of diameter of riser pipe, heater and the height of the riser on this instability are also investigated. (orig.) [de

  17. Study on flow instability for feasibility of a thin liquid film first wall

    Energy Technology Data Exchange (ETDEWEB)

    Okino, Fumito, E-mail: fumito.okino@iae.kyoto-u.ac.jp [Kyoto University Graduate School of Energy Science, Gokasho Uji, Kyoto (Japan); Kasada, Ryuta; Konishi, Satoshi [Kyoto University Institute of Advanced Energy, Gokasho Uji, Kyoto (Japan)

    2014-10-15

    Highlights: • We propose a probability of an instability wave growth on a liquid metal first wall. • Evaporated gas by the high energy flux is predicted to agitate this instability wave. • Liquid Pb-17Li with a velocity 10 m/s, the ambient gas must be below 6.2 × 10{sup 3} Pa. • This pressure corresponds to 1600 K and it is attainable under a fusion energy flux. • This probability is not yet verified so the full verifications are to be performed. - Abstract: This study proposes a probability of the evaporated gas that agitates a growing instability wave in a thin liquid film first wall. The liquid first wall was considered to be in vacuum and the effect of the ambient gas was neglected but the evaporated gas by the high energy fluxes is a probable cause of unstable wave agitation. The criterion is approximately expressed by the density ratio (Q{sub 2}) and the Weber number (We) as Q{sub 2} × We{sup 0.5} ≈ 5 × 10{sup −4}. Performed indirect experimental supported this criterion. For a case study of liquid Pb-17Li film with a velocity of 10 m/s, the evaporated gas pressure must be below 6.2 × 10{sup 3} Pa to maintain stable conditions. By recent study, this pressure is generated at 1600 K temperature and it is believed to be attainable by the energy fluxes on the first wall. This result is so far not confirmed so the full verification by experimental is to be performed.

  18. Studies of energetic-ion-driven MHD instabilities in helical plasmas with low magnetic shear

    International Nuclear Information System (INIS)

    Yamamoto, S.; Ascasibar, E.; Jimenez-Gomez, R.

    2012-11-01

    We discuss the features of energetic-ion-driven MHD instabilities such as Alfvén eigenmodes (AEs) in three-dimensional magnetic configuration with low magnetic shear and low toroidal field period number (N p ) that are characteristic of advanced helical plasmas. Comparison of experimental and numerical studies in Heliotron J with those in TJ-II indicates that the most unstable AE is global AE (GAE) in low magnetic shear configuration in spite of the iota and the helicity-induced AE (HAE) is also the most unstable AE in the high iota configuration. (author)

  19. Hard X-ray photoelectron spectroscopy study for transport behavior of CsI in heating test simulating a BWR severe accident condition: Chemical effects of boron vapors

    Energy Technology Data Exchange (ETDEWEB)

    Okane, T., E-mail: okanet@spring8.or.jp [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Kobata, M. [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Sato, I. [Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Kobayashi, K. [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Osaka, M. [Nuclear Science and Engineering Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Yamagami, H. [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Faculty of Science, Kyoto Sangyo University, Motoyama, Kamigamo, Kita-ku, Kyoto, 603-8555 (Japan)

    2016-02-15

    Highlights: • We have clarified the temperature-dependent chemical forms of Cs/I products. • We have examined the CsI-decomposing effects of B{sub 2}O{sub 3} vapor. • The possibility of Cs re-evaporation from CsI-deposited surface is suggested. • We have demonstrated the usefulness of HAXPES on FP chemistry. - Abstract: Transport behavior of CsI in the heating test, which simulated a BWR severe accident, was investigated by hard X-ray photoelectron spectroscopy (HAXPES) with an emphasis on the chemical effect of boron vapors. CsI deposited on metal tube at temperatures ranging from 150 °C to 750 °C was reacted with vapor/aerosol B{sub 2}O{sub 3}, and the chemical form of reaction products on the sample surface was examined from the HAXPES spectra of core levels, e.g., Ni 2p, Cs 3d and I 3d levels, and valence band. For the samples at ∼300 °C, while the chemical form of major product on the sample surface without an exposure to B{sub 2}O{sub 3} was suggested to be CsI from the HAXPES spectra, an intensity ratio of Cs/I was dramatically reduced at the sample surface after the reaction with B{sub 2}O{sub 3}. The results suggest the possibility of significant decomposition of deposited CsI induced by the chemical reaction with B{sub 2}O{sub 3} at specific temperatures.

  20. [Study on genetic instability of nm23H1 gene in Chinese with original gallbladder tumor].

    Science.gov (United States)

    Lu, Hai Ying; Zhang, Guo Qiang; Li, Ji Cheng

    2006-06-01

    The aim of this study was to examine the microsatellite instability (MSI) and loss of heterozygosity (LOH) of locus D17S396 on chromosome 17 and their influence on the expression of nm23H1 in gallbladder tumors, which may provide experimental basis for the tumor occurrence and metastasis. Techniques such as DNA extraction from formalin-fixed paraffin-embedded tissues, polymerase chain reaction-single strand conformation polymorphism (PCR-SSCP), ordinary silver stain were used to study MSI and LOH of locus D17S396. Envision immunohistochemistry and Leica-Qwin computer imaging techniques were used to assess the expression of gene nm23H1. In our experiment, the frequency of genetic instability of malignant gallbladder tumors was 42.55%, which was higher than that of gallbladder adenomas, while there were no genetic instability occurred in chronic cholecystitis tissue. The frequency of LOH seemed higher with the deteriorism of gallbladder tumor. Among 47 gallbladder carcinomas, the frequency of LOH and MSI were different between different differentiation cases (P gallbladder carcinoma, gallbladder adenoma and chronic cholecystitis tissue were different (P gallbladder carcinomas, the positive frequency of nm23H1 protein in LOH positive group was lower than that of LOH negative group (P gallbladder tumor. Both MSI and LOH of nm23H1 gene controlled the development of gallbladder tumor independently in different paths. MSI may be an early stage molecule marker of gallbladder carcinoma. LOH may be molecule marker for the deteriorism of gallbladder tissue, which could inhibit the expression of nm23H1 in local tissue of gallbladder carcinoma and endow it with high aggressive and poor prognosis. Increasing the amount of nm23H1 protein expression could effectively restrain gallbladder carcinoma metastasis and improve prognosis of patients.

  1. Isolated and combined medial patellofemoral ligament reconstruction in revision surgery for patellofemoral instability: a prospective study.

    Science.gov (United States)

    Kohn, Ludwig M; Meidinger, Gebhart; Beitzel, Knut; Banke, Ingo J; Hensler, Daniel; Imhoff, Andreas B; Schöttle, Philip B

    2013-09-01

    Persistent pain and redislocations after surgical treatment of patellofemoral instability are described in up to 40% of patients. However, prospective outcome data about revision surgery are missing. To evaluate the clinical outcome after revision medial patellofemoral ligament (MPFL) reconstruction using isolated and combined procedures, with a follow-up of 24 months. Case series; Level of evidence, 4. Study participants were 42 patients (median age, 22 years; range, 13-46 years) who underwent revision surgery between January 2007 and December 2009 because of persistent patellofemoral instability after a mean of 1.8 previous failed surgical interventions (lateral release, medial imbrication/vastus medialis obliquus distalization, medialization of the tuberosity). An isolated MPFL reconstruction was performed in 15 cases, while a combination procedure was performed in 27 cases. The clinical results were evaluated preoperatively and 24 months postoperatively using the International Knee Documentation Committee (IKDC), Kujala, and Tegner scores as well as a subjective questionnaire. Patellar shift, tilt, and height, as well as level of degeneration, were defined preoperatively and at the latest follow-up on plain radiographs and magnetic resonance imaging. At 24-month follow-up, 87% of the patients were satisfied or very satisfied with the treatment. No apprehension or redislocation was reported at follow-up, and there was a significant decrease in pain during daily activities. There were significant improvements (P instability is a multifactorial problem, revision surgery should be indicated only after a comprehensive examination. The results of this study show that MPFL reconstruction, alone or in combination, seems to be an effective treatment for recurrent patellar dislocations after a failed previous surgery, leading to significant increases in stability and functionality as well as a reduction in pain.

  2. The Posterior Bundle's Effect on Posteromedial Elbow Instability After a Transverse Coronoid Fracture: A Biomechanical Study.

    Science.gov (United States)

    Shukla, Dave R; Golan, Elan; Weiser, Mitch C; Nasser, Philip; Choueka, Jack; Hausman, Michael

    2018-04-01

    There has been increased interest in the role of the posterior bundle of the medial collateral ligament (pMUCL) in the elbow, particularly its effects on posteromedial rotatory stability. The ligament's effect in the context of an unfixable coronoid fracture has not been the focus of any study. The purposes of this biomechanical study were to evaluate the stabilizing effect of the pMUCL with a transverse coronoid fracture and to assess the effect of graft reconstruction of the ligament. We simulated a varus and internal rotatory subluxation in 7 cadaveric elbows at 30°, 60°, and 90° elbow flexion. The amount of ulnar rotation and medial ulnohumeral joint gapping were assessed in the intact elbow after we created a transverse coronoid injury, after we divided the pMUCL, and finally, after we performed a graft reconstruction of the pMUCL. At all angles tested, some stability was lost after cutting the pMUCL once the coronoid had been injured, because mean proximal ulnohumeral joint gapping increased afterward by 2.1, 2.2, and 1.3 mm at 90°, 60°, and 30°, respectively. Ulnar internal rotation significantly increased after pMUCL transection at 90°. At 60° and 30° elbow flexion, ulnar rotation increased after resection of the coronoid but not after pMUCL resection. An uninjured pMUCL stabilizes against varus internal rotatory instability in the setting of a transverse coronoid fracture at higher flexion angles. Further research is needed to optimize graft reconstruction of the pMUCL. The pMUCL is an important secondary stabilizer against posteromedial instability in the coronoid-deficient elbow. In the setting of an unfixable coronoid fracture, the surgeon should examine for posteromedial instability and consider addressing the pMUCL surgically. Copyright © 2018 American Society for Surgery of the Hand. Published by Elsevier Inc. All rights reserved.

  3. An experimental study on the flow instabilities and critical heat flux under natural circulation

    International Nuclear Information System (INIS)

    Kim, Yun Il

    1993-02-01

    This study has been carried out to investigate the hydrodynamic stabilities of natural circulation and to analyze Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations and the average mass flux have been observed. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. The results show that the flow instability has strongly dependent on geometric conditions and operating parameters, the inlet temperature and the heat flux of test section. It was found that unstable region for the heat flux and the inlet temperature exists between the single-phase stable region of low heat and low inlet temperature and the two-phase stable region of very high heat flux and high inlet temperature. The CHF data from the natural and forced circulation experiments have been compared each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions were low flow less than 70 kg/m 2 s of water in vertical round tube with diameter of 0.008m at near atmospheric pressure. In this study, no difference in CHF values is observed between natural and fored circulation. Since low flow usually has the oscillation characteristic of relatively low amplitude and high frequency, the effect of the flow instabilities on the CHF seems to be negligible

  4. The impact of BWR MK I primary containment failure dynamics on secondary containment integrity

    International Nuclear Information System (INIS)

    Greene, S.R.

    1987-01-01

    During the past four years, the ORNL BWRSAT Program has developed a series of increasingly sophisticated BWR secondary containment models. These models have been applied in a variety of studies to evaluate the severe accident mitigation capability of BWR secondary containments. This paper describes the results of a recent ORNL study of the impact of BWR MK I primary containment failure dynamics on secondary containment integrity. A 26-cell MELCOR Browns Ferry secondary containment model is described and the predicted thermodynamic response of the secondary containment to a variety of postulated primary containment failure modes is presented. The effects of primary containment failure location, timing, and ultimate hole size on secondary containment response is investigated, and the potential impact of hydrogen deflagrations on secondary containment integrity is explored

  5. 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)

  6. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  7. Delivering high performance BWR fuel reliably

    International Nuclear Information System (INIS)

    Schardt, J.F.

    1998-01-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  8. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  9. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  10. Study of effect of a smooth hump on hypersonic boundary layer instability

    Science.gov (United States)

    Park, Donghun; Park, Seung O.

    2016-12-01

    Effect of a two-dimensional smooth hump on linear instability of hypersonic boundary layer is studied by using parabolized stability equations. Linear evolution of mode S over a hump is analyzed for Mach 4.5 and 5.92 flat plate and Mach 7.1 sharp cone boundary layers. Mean flow for stability analysis is obtained by solving the parabolized Navier-Stokes equations. Hump with height smaller than local boundary layer thickness is considered. The case of flat plate and sharp cone without the hump are also studied to provide comparable data. For flat plate boundary layers, destabilization and stabilization effect is confirmed for hump located at upstream and downstream of synchronization point, respectively. Results of parametric studies to examine the effect of hump height, location, etc., are also given. For sharp cone boundary layer, stabilization influence of hump is also identified for a specific range of frequency. Stabilization influence of hump on convective instability of mode S is found to be a possible cause of previous experimental observations of delaying transition in hypersonic boundary layers.

  11. Studies of bandwidth dependence of laser plasma instabilities driven by the Nike laser

    Science.gov (United States)

    Weaver, J.; Kehne, D.; Obenschain, S.; Serlin, V.; Schmitt, A. J.; Oh, J.; Lehmberg, R. H.; Brown, C. M.; Seely, J.; Feldman, U.

    2012-10-01

    Experiments at the Nike laser facility of the Naval Research Laboratory are exploring the influence of laser bandwidth on laser plasma instabilities (LPI) driven by a deep ultraviolet pump (248 nm) that incorporates beam smoothing by induced spatial incoherence (ISI). In early ISI studies with longer wavelength Nd:glass lasers (1054 nm and 527 nm),footnotetextObenschain, PRL 62(1989);Mostovych, PRL 62(1987);Peyser, Phys. Fluids B 3(1991). stimulated Raman scattering, stimulated Brillouin scattering, and the two plasmon decay instability were reduced when wide bandwidth ISI (δν/ν˜0.03-0.19%) pulses irradiated targets at moderate to high intensities (10^14-10^15 W/cm^2). The current studies will compare the emission signatures of LPI from planar CH targets during Nike operation at large bandwidth (δν˜1THz) to observations for narrower bandwidth operation (δν˜0.1-0.3THz). These studies will help clarify the relative importance of the short wavelength and wide bandwidth to the increased LPI intensity thresholds observed at Nike. New pulse shapes are being used to generate plasmas with larger electron density scale-lengths that are closer to conditions during pellet implosions for direct drive inertial confinement fusion.

  12. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

    Greene, S.R.

    1986-01-01

    The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

  13. Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

    2009-01-01

    In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2007, TENPES published a revised edition of the guideline. This is the report of the revised edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent accumulation of radiolysis gas. (author)

  14. 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.

  15. 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

    International Nuclear Information System (INIS)

    Flores S, V. H.

    2011-01-01

    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 2 Pt (OH) 6 and Na 3 Rh (NO 2 ) 6 , Silver (Ag) with an aqueous solution of AgNO 3 , zirconium (Zr) with aqueous Zr O (NO 3 ) and ZrO 2 , and zinc (Zn) in aqueous solution of Zn (NO 3 ) 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 formed on the surface of 304l stainless steel

  16. Negative capacitance and instability at electrified interfaces: Lessons from the study of membrane capacitors

    Directory of Open Access Journals (Sweden)

    M.B.Partenskii

    2005-01-01

    Full Text Available Various models leading to predictions of negative capacitance, C, are briefly reviewed. Their relation to the nature of electric control is discussed. We reconfirm that the calculated double layer capacitance can be negative under σ-control - an artificial construct that requires uniform distribution of the electrode surface charge density, σ. However, only the total charge C (or the average surface charge density can be experimentally fixed in isolated cell studies (q-control. For those σ where C becomes negative under σ-control, the transition to q-control (i.e. relaxing the lateral change density distribution, fixing its mean value to σ leads to instability of the uniform distribution and a transition to a non-uniform phase. As an illustration, a “membrane capacitor” model is discussed. This exactly solvable model, allowing for both uniform and inhomogeneous relaxation of the electrical double layer, helps to demonstrate both the onset and some important features of the instability. Possibilities for further development are discussed briefly.

  17. Algorithm and exploratory study of the Hall MHD Rayleigh-Taylor instability

    International Nuclear Information System (INIS)

    Gardiner, Thomas Anthony

    2010-01-01

    This report is concerned with the influence of the Hall term on the nonlinear evolution of the Rayleigh-Taylor (RT) instability. This begins with a review of the magnetohydrodynamic (MHD) equations including the Hall term and the wave modes which are present in the system on time scales short enough that the plasma can be approximated as being stationary. In this limit one obtains what are known as the electron MHD (EMHD) equations which support two characteristic wave modes known as the whistler and Hall drift modes. Each of these modes is considered in some detail in order to draw attention to their key features. This analysis also serves to provide a background for testing the numerical algorithms used in this work. The numerical methods are briefly described and the EMHD solver is then tested for the evolution of whistler and Hall drift modes. These methods are then applied to study the nonlinear evolution of the MHD RT instability with and without the Hall term for two different configurations. The influence of the Hall term on the mixing and bubble growth rate are analyzed.

  18. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

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

    1991-01-01

    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)

  19. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  20. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    International Nuclear Information System (INIS)

    Aounallah, Y.

    2001-01-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  1. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

  2. Study on mechanism of combustion instability in a dump gas turbine combustor

    International Nuclear Information System (INIS)

    Lee, Yeon Joo; Lee, Jong Ho; Jeon, Chong Hwan; Chang, Yonng June

    2002-01-01

    Combustion instabilities are an important concern associated with lean premixed combustion. Laboratory-scale dump combustor was used to understand the underlying mechanisms causing combustion instabilities. Experiments were conducted at atmospheric pressure and sound level meter was used to track the pressure fluctuations inside the combustor. Instability maps and phase-resolved OH chemiluminescence images were obtained at several conditions to investigate the mechanism of combustion instability and relations between pressure wave and heat release rate. It showed that combustion instability was susceptible to occur at higher value of equivalence ratio (>0.6) as the mean velocity was decreased. Instabilities exhibited a longitudinal mode with a dominant frequency of ∼341.8 Hz, which corresponded to a quarter wave mode of combustor. Heat release and pressure waves were in-phase when instabilities occurred. Rayleigh index distribution gave a hint about the location where the strong coherence of pressure and heat release existed. These results also give an insight to the control scheme of combustion instabilities. Emission test revealed that NO x emissions were affected by not only equivalence but also combustion instability

  3. Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

    International Nuclear Information System (INIS)

    Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

    1992-01-01

    The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

  4. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (author)

  5. Longer reaction time of the fibularis longus muscle and reduced postural control in basketball players with functional ankle instability: A pilot study.

    Science.gov (United States)

    Méndez-Rebolledo, Guillermo; Guzmán-Muñoz, Eduardo; Gatica-Rojas, Valeska; Zbinden-Foncea, Hermann

    2015-08-01

    Motor control evaluation in subjects with functional ankle instability is questionable when both ankles of the same subject are compared (affected vs non-affected). To compare the postural control and reaction time of ankle muscles among: basketball players with FAI (instability group), basketball players without FAI (non-instability group) and healthy non-basketball-playing participants (control group). Case-control study. Laboratory. Instability (n = 10), non-instability (n = 10), and control groups (n = 11). Centre of pressure variables (area, velocity and sway) were measured with a force platform. Reaction time of ankle muscles was measured via electromyography. A one-way ANOVA demonstrated that there were significant differences between the instability and non-instability groups in the fibularis longus (p postural control and longer reaction time of the fibularis and tibialis anterior muscles. Copyright © 2014 Elsevier Ltd. All rights reserved.

  6. 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)

  7. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  8. Tracking Code for Microwave Instability

    International Nuclear Information System (INIS)

    Heifets, S.; SLAC

    2006-01-01

    To study microwave instability the tracking code is developed. For bench marking, results are compared with Oide-Yokoya results [1] for broad-band Q = 1 impedance. Results hint to two possible mechanisms determining the threshold of instability

  9. Study of neutronic problems related to the xenon instability in the pressurized water reactor

    International Nuclear Information System (INIS)

    Mathonniere, G.

    1988-03-01

    Xenon instabilities lead to increase initial errors. So it is extremely difficult to get an accurate calculation schema able to deal with these problems which are very important for electrical grid whose electronuclear output is large. The main task was to build a tridimensional calculation schema including nuclear, thermal hydraulic feedback. It was qualified by comparing its results with an actual experiment. Many technical problems were investigated through analytical studies or calculation results: time step, spatial mesh, finite element, convergence criterion, modelization for grids and fast control rod moves. Moreover, a 1D model, for less expensive, was studied. Its comparison with the 3D results allowed to check its accuracy and to validate the modelization of the radial laplacien and the modelization of the control rods [fr

  10. Study on parallel-channel asymmetry in supercritical flow instability experiment

    International Nuclear Information System (INIS)

    Xiong Ting; Yu Junchong; Yan Xiao; Huang Yanping; Xiao Zejun; Huang Shanfang

    2013-01-01

    Due to the urgent need for experimental study on supercritical water flow instability, the parallel-channel asymmetry which determines the feasibility of such experiments was studied with the experimental and numerical results in parallel dual channel. The evolution of flow rates in the experiments was analyzed, and the steady-state characteristics as well as transient characteristics of the system were obtained by self-developed numerical code. The results show that the asymmetry of the parallel dual channel would reduce the feasibility of experiments. The asymmetry of flow rates is aroused by geometrical asymmetry. Due to the property variation characteristics of supercritical water, the flow rate asymmetry is enlarged while rising beyond the pseudo critical point. The extent of flow rate asymmetry is affected by the bulk temperature and total flow rate; therefore the experimental feasibility can be enhanced by reducing the total flow rate. (authors)

  11. 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)

  12. Lateral movements in Rayleigh-Taylor instabilities due to frontiers. Experimental study

    Science.gov (United States)

    Binda, L.; Fernández, D.; El Hasi, C.; Zalts, A.; D'Onofrio, A.

    2018-01-01

    Lateral movements of the fingers in Rayleigh-Taylor hydrodynamic instabilities at the interface between two fluids are studied. We show that transverse movements appear when a physical boundary is present; these phenomena have not been explained until now. The boundary prevents one of the fluids from crossing it. Such frontiers can be buoyancy driven as, for example, the frontier to the passage of a less dense solution through a denser solution or when different aggregation states coexist (liquid and gaseous phases). An experimental study of the lateral movement velocity of the fingers was performed for different Rayleigh numbers (Ra), and when oscillations were detected, their amplitudes were studied. Liquid-liquid (L-L) and gas-liquid (G-L) systems were analysed. Aqueous HCl and Bromocresol Green (sodium salt, NaBCG) solutions were used in L-L experiments, and CO2 (gas) and aqueous NaOH, NaHCO3, and CaCl2 solutions were employed for the G-L studies. We observed that the lateral movement of the fingers and finger collapses near the interface are more notorious when Ra increases. The consequences of this, for each experience, are a decrease in the number of fingers and an increase in the velocity of the lateral finger movement close to the interface as time evolves. We found that the amplitude of the oscillations did not vary significantly within the considered Ra range. These results have an important implication when determining the wave number of instabilities in an evolving system. The wave number could be strongly diminished if there is a boundary.

  13. Subatmospheric boiling study of the operation of a horizontal thermosyphon reboiler loop: Instability

    International Nuclear Information System (INIS)

    Agunlejika, Ezekiel O.; Langston, Paul A.; Azzopardi, Barry J.; Hewakandamby, Buddhika N.

    2016-01-01

    Graphical abstract: The highlight of the characteristics of the geysering instability from analysed WMS data. Pictorial view of geysering instability, heat flux 9 kW/m"2 (P_S = 1.14 bar(a)), Static head = 1.265 m, valve setting = 1.0, process side pressure = 0.5 bar(a). - Highlights: • Characteristics of geysering instability in a horizontal thermosyphon reboiler loop is highlighted using Wire Mesh Sensor. • Interconnection between geysering instability and accompanying churn flow is identified. • Effects of stability parameters and pressure drop feedbacks on the loop at low heat fluxes are described. - Abstract: Distillation and chemical processing under vacuum is of immense interest to petroleum and chemical industries due to lower energy costs and improved safety. To tap into these benefits, energy efficient reboilers with lower maintenance costs are required. Here, a horizontal thermosyphon reboiler is investigated at subatmospheric pressures and low heat fluxes. This paper presents detailed experimental data obtained using Wire Mesh Sensor in a gas-liquid flow with heat transfer as well as temperatures, pressures and recirculation rates around the loop. Flow regimes which have been previously identified in other systems were detected. The nature of the instability which underpins the mechanisms involved and conditions aiding instability are reported. Churn flow pattern is persistently detected during instability. The nature of the instability and existence of oscillatory churn flow are interconnected.

  14. A numerical study of lowest-order short-crested water wave instabilities

    DEFF Research Database (Denmark)

    Fuhrman, David R.; Madsen, Per A.

    2005-01-01

    This work presents the first numerical simulations of the long-term evolution of doubly-periodic short-crested wave instabilities, which are the simplest cases involving the three-dimensional instability of genuinely three-dimensional progressive water waves. The simulated evolutions reveal quali...

  15. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

  16. Study on Roll Instability Mechanism and Stability Index of Articulated Steering Vehicles

    Directory of Open Access Journals (Sweden)

    Xuefei Li

    2016-01-01

    Full Text Available This study examines the roll instability mechanism and stability index of articulated steering vehicles (ASVs by taking wheel loaders as the research object. A seven-degree-of-freedom nonlinear dynamics model of the ASVs is built on the basis of multibody dynamics. A physical prototype model of an ASV is designed and manufactured to validate the dynamic model. Test results reasonably agree with the simulation results, which indicates that the established dynamic model can reasonably describe ASV movements. Detailed analysis of the rollover stability of the wheel loader is performed with the use of the established dynamic model. Analysis results show that rollover will occur when the roll angular velocity exceeds a critical threshold, which is affected by lateral acceleration and slope angle. On this basis, a dynamic stability index applicable to the ASVs is presented.

  17. Experimental study of the reversible behavior of modulational instability in optical fibers

    Science.gov (United States)

    van Simaeys, Gaetan; Emplit, Philippe; Haelterman, Marc

    2002-03-01

    We report what is to our knowledge the first clear-cut experimental evidence of the reversibility of modulational instability in dispersive Kerr media. It was possible to perform this experiment with standard telecommunication fiber because we used a specially designed 550-ps square-pulse laser source based on the two-wavelength configuration of a nonlinear optical loop mirror. Our observations demonstrate that reversibility is due to well-balanced and synchronous energy transfer among a significant number of spectral wave components. These results provide what we believe is the first evidence, in the field of nonlinear optics, of the universal Fermi-Pasta-Ulam recurrence phenomenon that has been predicted for a large number of conservative nonlinear systems, including those described by a nonlinear Schrödinger equation that is relevant to the context of the present study.

  18. Study of an instability of the PEP-II positron beam (Ohmi effect and multipactoring)

    International Nuclear Information System (INIS)

    Heifets, S.A.

    1995-11-01

    The paper is organized in the following way. First, Ohmi effect induced by direct flow of primary photoelectrons is studied for the PEP-II parameters. The production rate and kinematics take into account the antechamber of the LER. We discuss the effect of the secondary emission of electrons in the AL chamber, where the yield is larger than one. Resonance multipactoring is considered, and then the average density of the secondary electrons is estimated taking into account the space-charge effect and the interaction with the beam. We show that in the extreme case there is a self-consistent regime similar to the regime of the space-charge dominated cathode. Finally, the rate of ion production by accumulated electrons and the possibility of the ion induced pressure instability is discussed

  19. Back-Analyses of Landfill Instability Induced by High Water Level: Case Study of Shenzhen Landfill

    Directory of Open Access Journals (Sweden)

    Ren Peng

    2016-01-01

    Full Text Available In June 2008, the Shenzhen landfill slope failed. This case is used as an example to study the deformation characteristics and failure mode of a slope induced by high water levels. An integrated monitoring system, including water level gauges, electronic total stations, and inclinometers, was used to monitor the slope failure process. The field measurements suggest that the landfill landslide was caused by a deep slip along the weak interface of the composite liner system at the base of the landfill. The high water level is considered to be the main factor that caused this failure. To calculate the relative interface shear displacements in the geosynthetic multilayer liner system, a series of numerical direct shear tests were carried out. Based on the numerical results, the composite lining system simplified and the centrifuge modeling technique was used to quantitatively evaluate the effect of water levels on landfill instability.

  20. Back-Analyses of Landfill Instability Induced by High Water Level: Case Study of Shenzhen Landfill

    Science.gov (United States)

    Peng, Ren; Hou, Yujing; Zhan, Liangtong; Yao, Yangping

    2016-01-01

    In June 2008, the Shenzhen landfill slope failed. This case is used as an example to study the deformation characteristics and failure mode of a slope induced by high water levels. An integrated monitoring system, including water level gauges, electronic total stations, and inclinometers, was used to monitor the slope failure process. The field measurements suggest that the landfill landslide was caused by a deep slip along the weak interface of the composite liner system at the base of the landfill. The high water level is considered to be the main factor that caused this failure. To calculate the relative interface shear displacements in the geosynthetic multilayer liner system, a series of numerical direct shear tests were carried out. Based on the numerical results, the composite lining system simplified and the centrifuge modeling technique was used to quantitatively evaluate the effect of water levels on landfill instability. PMID:26771627

  1. Experimental and numerical study of plastic shear instability under high-speed loading conditions

    International Nuclear Information System (INIS)

    Sokovikov, Mikhail; Chudinov, Vasiliy; Bilalov, Dmitry; Oborin, Vladimir; Uvarov, Sergey; Plekhov, Oleg; Terekhina, Alena; Naimark, Oleg

    2014-01-01

    The behavior of specimens dynamically loaded during the split Hopkinson (Kolsky) bar tests in a regime close to simple shear conditions was studied. The lateral surface of the specimens was investigated in a real-time mode with the aid of a high-speed infra-red camera CEDIP Silver 450M. The temperature field distribution obtained at different time made it possible to trace the evolution of plastic strain localization. The process of target perforation involving plug formation and ejection was examined using a high-speed infra-red camera and a VISAR velocity measurement system. The microstructure of tested specimens was analyzed using an optical interferometer-profilometer and a scanning electron microscope. The development of plastic shear instability regions has been simulated numerically

  2. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  3. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

    1986-01-01

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  4. BWR mechanics and materials technology update

    International Nuclear Information System (INIS)

    Kiss, E.

    1983-01-01

    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. (orig.)

  5. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Suzuki, H.; Hatamiya, S.; Murase, M.

    1988-01-01

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

  6. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

  7. 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.

  8. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1992-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

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

    International Nuclear Information System (INIS)

    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

  10. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  11. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  12. Delivering high performance BWR fuel reliably

    Energy Technology Data Exchange (ETDEWEB)

    Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

    1998-07-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  13. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  14. PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

    OpenAIRE

    MELARA SAN ROMÁN, JOSÉ

    2016-01-01

    [EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

  15. Numerical study of the effect of inlet geometry on combustion instabilities in a lean premixed swirl combustor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Eon [Dept. of Mechanical Engineering, Inha University, Incheon (Korea, Republic of); Park, Seul Hyun [Dept. of Mechanical Systems Engineering, Chosun University, Gwangju (Korea, Republic of); Hwang, Cheol Hong [Dept. of Fire and Disaster Prevention, Daejeon University, Daejeon (Korea, Republic of)

    2016-11-15

    The effects of flow structure and flame dynamics on combustion instabilities in a lean premixed swirl combustor were numerically investigated using Large eddy simulation (LES) by varying the inlet geometry of combustor. The dynamic ksgs-equation and G-equation flamelet models were respectively employed as the LES subgrid models of turbulence and combustion. The divergent half angle (α) in the combustor inlet was varied systematically from 30° to 90° to quantify the effect of inlet geometry on the combustion instabilities. This variation caused considerable deformation in recirculation zones in terms of their size and location, leading to significant changes in flame dynamics. Analysis of unsteady pressure distributions in the combustor showed that the largest damping caused by combustion instabilities takes place at α = 45°, and the amplitude of acoustic pressure oscillation is largest at α = 30°. Examination of local Rayleigh parameters indicated that controlling flame-vortex interactions by modifying inlet geometry can change the local characteristics of combustion instabilities in terms of their amplification and suppression, and thus serve as a useful approach to reduce the instabilities in a lean premixed swirl combustor. These phenomena were studied in detail through unsteady analysis associated with flow and flame dynamics.

  16. Numerical study of the effect of inlet geometry on combustion instabilities in a lean premixed swirl combustor

    International Nuclear Information System (INIS)

    Lee, Chang Eon; Park, Seul Hyun; Hwang, Cheol Hong

    2016-01-01

    The effects of flow structure and flame dynamics on combustion instabilities in a lean premixed swirl combustor were numerically investigated using Large eddy simulation (LES) by varying the inlet geometry of combustor. The dynamic ksgs-equation and G-equation flamelet models were respectively employed as the LES subgrid models of turbulence and combustion. The divergent half angle (α) in the combustor inlet was varied systematically from 30° to 90° to quantify the effect of inlet geometry on the combustion instabilities. This variation caused considerable deformation in recirculation zones in terms of their size and location, leading to significant changes in flame dynamics. Analysis of unsteady pressure distributions in the combustor showed that the largest damping caused by combustion instabilities takes place at α = 45°, and the amplitude of acoustic pressure oscillation is largest at α = 30°. Examination of local Rayleigh parameters indicated that controlling flame-vortex interactions by modifying inlet geometry can change the local characteristics of combustion instabilities in terms of their amplification and suppression, and thus serve as a useful approach to reduce the instabilities in a lean premixed swirl combustor. These phenomena were studied in detail through unsteady analysis associated with flow and flame dynamics

  17. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  18. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  19. Economic analysis of hydride fueled BWR

    International Nuclear Information System (INIS)

    Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

    2009-01-01

    The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

  20. Investigation of BWR stability in Forsmark 2

    International Nuclear Information System (INIS)

    Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

    1988-01-01

    A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  1. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  2. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  3. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1982-01-01

    A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  4. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.; Farmer, W.S.

    1992-01-01

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  5. Safety evaluation of BWR off-gas treatment systems

    International Nuclear Information System (INIS)

    Schultz, R.J.; Schmitt, R.C.

    1975-01-01

    Some of the results of a safety evaluation performed on current generic types of BWR off-gas treatment systems including cooled and ambient temperature adsorber beds and cryogenics are presented. The evaluation covered the four generic types of off-gas systems and the systems of five major vendors. This study was part of original work performed under AEC contract for the Directorate of Regulatory Standards. The analysis techniques employed for the safety evaluation of these systems include: Fault Tree Analysis; FMECA (Failure Mode Effects and Criticality Analysis); general system comparisons, contaminant, system control, and design adequacy evaluations; and resultant Off-Site Dose Calculations. The salient areas presented are some of the potential problem areas, the approach that industry has taken to mitigate or design against potential upset conditions, and areas where possible deficiencies still exist. Potential problem areas discussed include hydrogen detonation, hydrogen release to equipment areas, operator/automatic control interface, and needed engineering evaluation to insure safe system operation. Of the systems reviewed, most were in the category of advanced or improved over that commonly in use today, and a conclusion from the study was that these systems offer excellent potential for noble gas control for BWR power plants where more stringent controls may be specified -- now or in the future. (U.S.)

  6. Experimental study of static flow instability in subcooled flow boiling in parallel channels

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; McDuffee, J.L.; Yoder, G.L.

    1995-01-01

    Experimental data for static flow instability or flow excursion (FE) at conditions applicable to the Advanced Neutron Source Reactor are very limited. A series of FE tests with light water flowing vertically upward was completed covering a local exit heat flux range of 0.7--18 MW/m 2 , exit velocity range of 2.8--28.4 m/s, exit pressure range of 0.117--1.7 MPa, and inlet temperature range of 40-- 50 degrees C. Most of the tests were performed in a ''stiff'' (constant flow) system where the instability threshold was detected through the minimum of the pressure-drop curve. A few tests were also conducted using as ''soft'' (constant pressure drop) a system as possible to secure a true FE phenomenon (actual secondary burnout). True critical heat flux experiments under similar conditions were also conducted using a stiff system. The FE data reported in this study considerably extend the velocity range of data presently available worldwide, most of which were obtained at velocities below 10 m/s. The Saha and Zuber correlation had the best fit with the data out of the three correlations compared. However, a modification was necessary to take into account the demonstrated dependence of the St and Nu numbers on subcooling levels, especially in the low subcooling regime. Comparison of Thermal Hydraulic Test Loop (THTL) data, as well as extensive data from other investigators, led to a proposed modification to the Saha and Zuber correlation for onset of significant void, applied to FE prediction. The mean and standard deviation of the THTL data were 0.95 and 15%, respectively, when comparing the THTL data with the original Saha and Zuber correlation, and 0.93 and 10% when comparing them with the modification. Comparison with the worldwide database showed a mean and standard deviation of 1.37 and 53%, respectively, for the original Saha and Zuber correlation and 1.0 and 27% for the modification

  7. Analytical and numerical study of MHD instabilities development in magnetized accretion-ejection structures

    International Nuclear Information System (INIS)

    Kersale, Evy

    2000-01-01

    The first part of this work proposes a new version of the mathematical formalism used to describe pressure-driven instabilities in magnetized accretion-ejection structures. Such processes, occurring in magnetically confined plasmas, pose very stringent limits to thermonuclear fusion devices but their influence in astrophysical objects has rarely been considered. In a framework which eliminates fast magnetosonic waves one develops a system of equations allowing us to follow both ballooning and interchange modes. An application of this result to a cylindrical jet being subject to solid rotation shows that the inner parts of such structures are destabilized by magnetic shear. Furthermore, while clarifying somewhat previous studies, one finds that jets confined by a dominant toroidal magnetic field are generically unstable with respect to interchange modes. Moreover, one has written a numerical code to solve the MHD partial differential equations. Starting with a basic algorithm, one has assessed the effects of the geometry, boundary conditions and artificial dissipation on numerical computation. The code has been tested by solving classical hydrodynamic and MHD Riemann problems. A new mechanism of ultra high energy cosmic ray production in gamma-ray bursts composes the last part of this work. In these objects, particles are accelerated up to energies of the order of 10 21 eV, by means of relativistic Alfven perturbations crossings. A stream instability involving a highly relativistic shell of plasma, the fireball, and baryons going through it produces such Alfven fronts. Then, Brillouin-like backscattering processes redistribute the available energy between the forward and backward Alfven waves and the magnetosonic ones. (author) [fr

  8. Update on materials performance and electrochemistry in hydrogen water chemistry at Dresden-2 BWR

    International Nuclear Information System (INIS)

    Indig, M.E.; Weber, J.E.; Davis, R.B.; Gordon, B.M.

    1985-01-01

    Previous studies performed in 1982 indicated that if sufficient hydrogen was injected into the Dresden-2 BWR, IGSCC of sensitized austenitic stainless steel was mitigated. The present series of experiments were aimed at verification of the above finding, determining how much time off hydrogen water chemistry (HWC) could be tolerated and how HWC affected pre-existing cracks

  9. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  10. Nonlinear evolution of MHD instabilities

    International Nuclear Information System (INIS)

    Bateman, G.; Hicks, H.R.; Wooten, J.W.; Dory, R.A.

    1975-01-01

    A 3-D nonlinear MHD computer code was used to study the time evolution of internal instabilities. Velocity vortex cells are observed to persist into the nonlinear evolution. Pressure and density profiles convect around these cells for a weak localized instability, or convect into the wall for a strong instability. (U.S.)

  11. Simulation studies of plasma waves in the electron foreshock - The transition from reactive to kinetic instability

    Science.gov (United States)

    Dum, C. T.

    1990-01-01

    Particle simulation experiments were used to analyze the electron beam-plasma instability. It is shown that there is a transition from the reactive state of the electron beam-plasma instability to the kinetic instability of Langmuir waves. Quantitative tests, which include an evaluation of the dispersion relation for the evolving non-Maxwellian beam distribution, show that a quasi-linear theory describes the onset of this transition and applies again fully to the kinetic stage. This stage is practically identical to the late stage seen in simulations of plasma waves in the electron foreshock described by Dum (1990).

  12. BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

    International Nuclear Information System (INIS)

    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

  13. Study of three-dimensional Rayleigh--Taylor instability in compressible fluids through level set method and parallel computation

    International Nuclear Information System (INIS)

    Li, X.L.

    1993-01-01

    Computation of three-dimensional (3-D) Rayleigh--Taylor instability in compressible fluids is performed on a MIMD computer. A second-order TVD scheme is applied with a fully parallelized algorithm to the 3-D Euler equations. The computational program is implemented for a 3-D study of bubble evolution in the Rayleigh--Taylor instability with varying bubble aspect ratio and for large-scale simulation of a 3-D random fluid interface. The numerical solution is compared with the experimental results by Taylor

  14. Examination of minor actinide annihilation by BWR core

    International Nuclear Information System (INIS)

    Hida, Kazuki

    1995-01-01

    From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

  15. Transmutation of minor actinide using BWR fueled mixed oxide

    International Nuclear Information System (INIS)

    Susilo, Jati

    2000-01-01

    Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

  16. Evaluation of thermal margin during BWR neutron flux oscillation

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  17. A detailed BWR recirculation loop model for RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

    2017-01-15

    Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

  18. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    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 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  19. Kinetic study of the sausage mode of a resistive instability of a relativistic electron beam

    International Nuclear Information System (INIS)

    Gureev, K.G.; Zolotarev, V.O.; Stolbetsov, S.D.

    1984-01-01

    The nonlinear problem of the growth of the sausage mode of the resistive instability of a relativistic electron beam propagating without collisions through a tenuous plasma is solved. The plasma conductivity is assumed to be high, so that the wave phase velocity is low in comparison with the velocity of light. A kinetic approach is taken to the description of the beam. A numerical solution of the problem shows that this instability occurs in a cold, uniform beam. In the nonlinear stage of the instability the beam goes through states with a hollow structure. Suppression of the instability is found for a beam with a Bennett distribution function. The stabilization results from phase mixing of the beam particles

  20. Diagnosis of shoulder instability in dogs and cats: a retrospective study

    International Nuclear Information System (INIS)

    Bardet, J.F.

    1998-01-01

    The glenohumeral joint is a remarkable articulation providing the greatest range of motion of any joint in the body. Glenohumeral stabilityresults from several mechanisms, including those that do not require expenditure of energy by muscle ('''passive mechanisms'') and those that do (''active mechanisms''). Glenohumeral instability has been recognized in 47 shoulders of 45 dogs and one cat. Cases are presented because of chronic foreleg lameness. Shoulder joint pain is obviated by theorthopedic examination. Only 57% of the involved shoulders presented with degenerative joint disease. Signs of instability are recognized under anesthesia using a craniocaudal or mediolateral drawer sign or both. This report describes the radiographic and arthroscopic findings of shoulder instability. Arthroscopy of the shoulder joint allows identification of all intra-articular pathologies. Shoulder instability notfully recognized in the past, appears to be the most common cause of shoulder lameness in the dog

  1. Investigation of power oscillation mechanisms based on noise analysis at Forsmark-1 BWR

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1996-01-01

    Noise analysis has been performed for stability test data collected during reactor start-up in January 1989 at the boiling water reactor (BWR) Forsmark unit 1. A unique instrumentation to measure local coolant flow in this reactor allowed investigation of dynamic interactions between neutron flux and coolant flow noise signals at different radial positions in the core. The causal relationship for these signals was evaluated based on a method called signal transmission path (STP) analysis with the aim of identifying the principal mechanism of power oscillations in this reactor. The results of the present study indicated that large amplitude power oscillations were induced by two instability mechanisms concurrent in the core. The first is the global void reactivity feedback effect which played the most significant role to power oscillations at a resonant frequency of about 0.53 Hz. The second is the thermal-hydraulics coupling with neutron kinetics, inducing resonant oscillations at about 0.45 Hz. The latter was found to be active only in a certain core region. A peculiar phenomenon of amplitude modulations observed in some local power range monitor (LPRM) signals was also examined. It was interpreted to occur as the consequence of these two resonant power oscillations, the frequencies of which lie close to each other. The noise analysis technique applied in the present study is expected to be useful to get a deeper understanding of the power oscillation mechanism which is active in the reactor under evaluation. The technique may be applicable to BWRs with instruments to measure local channel flow together with in-core neutron detectors. (Author)

  2. Studies in the evolution of hydrodynamic instabilities and their role in inertial confinement fusion

    International Nuclear Information System (INIS)

    Shvarts, D.; Oron, D.; Sadot, O.

    2001-01-01

    Hydrodynamic instabilities, such as the Rayleigh-Taylor and Richtmyer-Meshkov instabilities, have a central role when trying to achieve net thermonuclear fusion energy via the method of Inertial Confinement Fusion. We shall review recent theoretical, numerical and experimental work that describes the evolution of two- and three-dimensional perturbations. Finally, the effects of these perturbation on the ignition conditions, using new self-similar solutions for perturbed burn wave propagation will be discussed. (author)

  3. Study of fracture and stress-induced morphological instabilities in polymeric materials

    Science.gov (United States)

    Sabouri-Ghomi, Mohsen

    We study the phenomena of fracture in polymers at the molecular and continuum level. At a molecular level, we study the failure of polymer/polymer interfaces. Our main focus is on a specific mode of failure known as chain pull-out fracture, which is common to weak adhesive junctions, and polymer blends and mixtures. In the case of the interface between incompatible polymers, reinforcement is achieved by adding a block copolymer to the interface. We introduce a microscopic model based on Brownian dynamics to investigate the effect of the polymerization index N, of the block connector chain, on fracture toughness of such reinforced polymeric junctions. We consider the mushroom regime, where connector chains are grafted with low surface density, for the case of large pulling velocity. We find that for short chains the interface fracture toughness depends linearly on the polymerization index N of the connector chains, while for longer chains the dependence becomes N 3/2. We propose a scaling argument, based on the geometry of the initial configuration, that accounts for both short and long chains and the crossover between them. At the continuum level, we study the pattern selection mechanism of finger-like crack growth phenomena in gradient driven growth problems in general, and the structure of stress-induced morphological instabilities in crazing of polymer glasses in particular. We simulate solidification in a narrow channel through the use of a phase-field model with an adaptive grid. By tuning a dimensionless parameter, the Peclet number, we show a continuous crossover from a free dendrite at high Peclet numbers to anisotropic viscous fingering at low Peclet numbers. At low Peclet numbers we find good agreement between our results, theoretical predictions, and experiment, providing the first quantitative test of solvability theory for anisotropic viscous fingers. For high undercoolings, we find new phenomena, a solid forger which satisfies stability and

  4. Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

    1996-01-01

    A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

  5. Lumbar segmental instability: a criterion-related validity study of manual therapy assessment

    Directory of Open Access Journals (Sweden)

    Chapple Cathy

    2005-11-01

    Full Text Available Abstract Background Musculoskeletal physiotherapists routinely assess lumbar segmental motion during the clinical examination of a patient with low back pain. The validity of manual assessment of segmental motion has not, however, been adequately investigated. Methods In this prospective, multi-centre, pragmatic, diagnostic validity study, 138 consecutive patients with recurrent or chronic low back pain (R/CLBP were recruited. Physiotherapists with post-graduate training in manual therapy performed passive accessory intervertebral motion tests (PAIVMs and passive physiological intervertebral motion tests (PPIVMs. Consenting patients were referred for flexion-extension radiographs. Sagittal angular rotation and sagittal translation of each lumbar spinal motion segment was measured from these radiographs, and compared to a reference range derived from a study of 30 asymptomatic volunteers. Motion beyond two standard deviations from the reference mean was considered diagnostic of rotational lumbar segmental instability (LSI and translational LSI. Accuracy and validity of the clinical assessments were expressed using sensitivity, specificity, and likelihood ratio statistics with 95% confidence intervals (CI. Results Only translation LSI was found to be significantly associated with R/CLBP (p Conclusion This study provides the first evidence reporting the concurrent validity of manual tests for the detection of abnormal sagittal planar motion. PAIVMs and PPIVMs are highly specific, but not sensitive, for the detection of translation LSI. Likelihood ratios resulting from positive test results were only moderate. This research indicates that manual clinical examination procedures have moderate validity for detecting segmental motion abnormality.

  6. Study of flow instabilities in double-channel natural circulation boiling systems

    International Nuclear Information System (INIS)

    Durga Prasad, Gonella V.; Pandey, Manmohan; Pradhan, Santosh K.; Gupta, Satish K.

    2008-01-01

    Natural circulation boiling systems consisting of parallel channels can undergo different types of oscillations (in-phase or out-of-phase) depending on the geometric parameters and operating conditions. Disturbances in one channel affect the flow in other channels, which triggers thermal-hydraulic oscillations. In the present work, the modes of oscillation under different operating conditions and channel-to-channel interaction during power fluctuations and on-power refueling in a double-channel natural circulation boiling system are investigated. The system is modeled using a lumped parameter mathematical model and RELAP5/MOD3.4. Parametric studies are carried out for an equal-power double-channel system, at different operating conditions, with both the models, and the results are compared. Instabilities, non-linear oscillations, and effects of recirculation loop dynamics and geometric parameters on the mode of oscillations, are studied using the lumped model. The two channels oscillate out-of-phase in Type-I region, but in Type-II region, both the modes of oscillation are observed under different conditions. Channel-to-channel interaction and on-power refueling studies are carried out using the RELAP model. At high powers, disturbances in one channel significantly affect the stability of the other channel. During on-power refueling, a near-stagnation condition or low-velocity reverse flow can occur, the possibility of reverse flow being higher at lower pressures

  7. The study of waves, instabilities, and turbulence using Thomson scattering in laser plasmas

    International Nuclear Information System (INIS)

    Drake, R.P.

    1995-01-01

    Much basic work in plasma physics has been devoted to the study of wave properties in plasmas, one of the nonlinear development of driven waves, and of the instabilities in which such waves may participate. The use of laser-plasma techniques has allowed one to extend such studies into new regimes. Such techniques and their results are the subject here. Once one chooses a physical problem within this subject area, it is now possible to design a laser-plasma experiment that is optimized for the study of that problem. The plasma can be designed to have a variety of density and flow-velocity profiles, the damping of ion acoustic waves and of electron plasma waves can be independently controlled, and the waves can be driven weakly or strongly. By using Nd-glass lasers and their harmonics one can non-invasively drive and diagnose the waves, using separate laser beams to produce the plasma, drive the waves, and diagnose their properties. The author uses as examples some recent work with his collaborators, including the first experimental detection of ion plasma waves and the first direct observation of the plasma wave driven by the acoustic decay of laser light

  8. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  9. GPE-BWR and the containment venting and filtering issue

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

    The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

  10. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  11. Identification of suitable reference genes for gene expression studies of shoulder instability.

    Directory of Open Access Journals (Sweden)

    Mariana Ferreira Leal

    Full Text Available Shoulder instability is a common shoulder injury, and patients present with plastic deformation of the glenohumeral capsule. Gene expression analysis may be a useful tool for increasing the general understanding of capsule deformation, and reverse-transcription quantitative polymerase chain reaction (RT-qPCR has become an effective method for such studies. Although RT-qPCR is highly sensitive and specific, it requires the use of suitable reference genes for data normalization to guarantee meaningful and reproducible results. In the present study, we evaluated the suitability of a set of reference genes using samples from the glenohumeral capsules of individuals with and without shoulder instability. We analyzed the expression of six commonly used reference genes (ACTB, B2M, GAPDH, HPRT1, TBP and TFRC in the antero-inferior, antero-superior and posterior portions of the glenohumeral capsules of cases and controls. The stability of the candidate reference gene expression was determined using four software packages: NormFinder, geNorm, BestKeeper and DataAssist. Overall, HPRT1 was the best single reference gene, and HPRT1 and B2M composed the best pair of reference genes from different analysis groups, including simultaneous analysis of all tissue samples. GenEx software was used to identify the optimal number of reference genes to be used for normalization and demonstrated that the accumulated standard deviation resulting from the use of 2 reference genes was similar to that resulting from the use of 3 or more reference genes. To identify the optimal combination of reference genes, we evaluated the expression of COL1A1. Although the use of different reference gene combinations yielded variable normalized quantities, the relative quantities within sample groups were similar and confirmed that no obvious differences were observed when using 2, 3 or 4 reference genes. Consequently, the use of 2 stable reference genes for normalization, especially

  12. Magnetohydrodynamic study of three-dimensional instability of the spontaneous fast magnetic reconnection

    International Nuclear Information System (INIS)

    Shimizu, T.; Kondoh, K.; Ugai, M.; Shibata, K.

    2009-01-01

    Three-dimensional instability of the spontaneous fast magnetic reconnection is studied with magnetohydrodynamic (MHD) simulation, where the two-dimensional model of the spontaneous fast magnetic reconnection is destabilized in three dimension. Generally, in two-dimensional magnetic reconnection models, every plasma condition is assumed to be uniform in the sheet current direction. In such two-dimensional MHD simulations, the current sheet destabilized by the initial resistive disturbance can be developed to fast magnetic reconnection by a current driven anomalous resistivity. In this paper, the initial resistive disturbance includes a small amount of fluctuations in the sheet current direction, i.e., along the magnetic neutral line. The other conditions are the same as that of previous two-dimensional MHD studies for fast magnetic reconnection. Accordingly, we may expect that approximately two-dimensional fast magnetic reconnection occurs in the MHD simulation. In fact, the fast magnetic reconnection activated on the first stage of the simulation is two dimensional. However, on the subsequent stages, it spontaneously becomes three dimensional and is strongly localized in the sheet current direction. The resulting three-dimensional fast magnetic reconnection intermittently ejects three-dimensional magnetic loops. Such intermittent ejections of the three-dimensional loops are similar to the intermittent downflows observed in the solar flares. The ejection of the three-dimensional loops seems to be random but, numerically and theoretically, it is shown that the aspect ratio of the ejected loops is limited under a criterion.

  13. Sternal instability measured with radiostereometric analysis. A study of method feasibility, accuracy and precision.

    Science.gov (United States)

    Vestergaard, Rikke Falsig; Søballe, Kjeld; Hasenkam, John Michael; Stilling, Maiken

    2018-05-18

    A small, but unstable, saw-gap may hinder bone-bridging and induce development of painful sternal dehiscence. We propose the use of Radiostereometric Analysis (RSA) for evaluation of sternal instability and present a method validation. Four bone analogs (phantoms) were sternotomized and tantalum beads were inserted in each half. The models were reunited with wire cerclage and placed in a radiolucent separation device. Stereoradiographs (n = 48) of the phantoms in 3 positions were recorded at 4 imposed separation points. The accuracy and precision was compared statistically and presented as translations along the 3 orthogonal axes. 7 sternotomized patients were evaluated for clinical RSA precision by double-examination stereoradiographs (n = 28). In the phantom study, we found no systematic error (p > 0.3) between the three phantom positions, and precision for evaluation of sternal separation was 0.02 mm. Phantom accuracy was mean 0.13 mm (SD 0.25). In the clinical study, we found a detection limit of 0.42 mm for sternal separation and of 2 mm for anterior-posterior dislocation of the sternal halves for the individual patient. RSA is a precise and low-dose image modality feasible for clinical evaluation of sternal stability in research. ClinicalTrials.gov Identifier: NCT02738437 , retrospectively registered.

  14. A numerical study of bubble interactions in Rayleigh--Taylor instability for compressible fluids

    International Nuclear Information System (INIS)

    Glimm, J.; Li, X.L.; Menikoff, R.; Sharp, D.H.; Zhang, Q.

    1990-01-01

    The late nonlinear and chaotic stage of Rayleigh--Taylor instability is characterized by the evolution of bubbles of the light fluid and spikes of the heavy fluid, each penetrating into the other phase. This paper is focused on the numerical study of bubble interactions and their effect on the statistical behavior and evolution of the bubble envelope. Compressible fluids described by the two-fluid Euler equations are considered and the front tracking method for numerical simulation of these equations is used. Two major phenomena are studied. One is the dynamics of the bubbles in a chaotic environment and the interaction among neighboring bubbles. Another one is the acceleration of the overall bubble envelope, which is a statistical consequence of the interactions of bubbles. The main result is a consistent analysis, at least in the approximately incompressible case of these two phenomena. The consistency encompasses the analysis of experiments, numerical simulation, simple theoretical models, and variation of parameters. Numerical simulation results that are in quantitative agreement with laboratory experiment for one-and-one-half (1 1/2) generations of bubble merger are presented. To the authors' knowledge, computations of this accuracy have not previously been obtained

  15. Study of the instability of black slags from electric arc furnace steel industry

    Directory of Open Access Journals (Sweden)

    Frías, M.

    2002-09-01

    Full Text Available In Spain, the steel manufacture produces important quantities of by-products, representing between 15 and 20 % of total steel production. Most by-products are deposited on open air spaces causing serious economical and environmental problems, internationally, different recycling wais are studied, being the main alternative for these by-products as recycled aggregate. The possibility of recycling these by-products in construction sector depends on its possible volume instability because of the presence of some undesirable compounds. In current paper, two different black slags from electric arc furnace steel industry were chemically characterized, paying attention to some well-known compounds by theirs expansion effects, such as: free CaO, free MgO, chlorides and sulphates. The analytical results carried out in the current research detected the presence of insignificant or null amounts of harmful compounds. Therefore, they should not have any negative incidence on phenomena of volume instability.

    En España la fabricación de acero produce grandes cantidades de residuos industriales, las cuales representan entre el 15-20 % de la producción total de acero, en su mayor parte se depositan en vertederos, causando serios problemas económicos y medioambientales a todos los sectores implicados. A nivel internacional, se están estudiando diferentes vías de reutilización, siendo su uso principal como árido de reciclado. La posibilidad de reutilizar estos subproductos industriales en el sector de la construcción se basa en su posible inestabilidad volumétrica, debido a la presencia de ciertos compuestos no deseados. En este trabajo se caracterizan químicamente 2 escorias negras de horno de arco eléctrico con diferente procedencia y se cuantifican algunos de los principales compuestos conocidos por sus efectos expansivos, como: cal libre, magnesia libre, cloruros y sulfatos. Los resultados analíticos de estas dos escorias negras muestran

  16. Prevention of organic iodide formation in BWR's

    International Nuclear Information System (INIS)

    Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

    1996-01-01

    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

  17. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

    1992-01-01

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  18. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

  19. An experimental study on the flow instabilities and critical heat flux under natural circulation

    International Nuclear Information System (INIS)

    Kim, Yun II; Chang, Soon Heung

    2004-01-01

    This study has been carried out to investigate the hydrodynamic stabilities and Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations has been observed and the average mass flux under flow oscillation have been measured. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. And the CHF data from low flow experiments, namely the natural and forced circulation, have been compared with each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions for the CHF experiments were a low flow of less than 70 kg/m 2 s of water in a vertical round tube with diameter of 0.008 m at near atmospheric pressure. (author)

  20. AFM-based tribological study of nanopatterned surfaces: the influence of contact area instabilities.

    Science.gov (United States)

    Rota, A; Serpini, E; Gazzadi, G C; Valeri, S

    2016-04-06

    Although the importance of morphology on the tribological properties of surfaces has long been proved, an exhaustive understanding of nanopatterning effects is still lacking due to the difficulty in both fabricating 'really nano-' structures and detecting their tribological properties. In the present work we show how the probe-surface contact area can be a critical parameter due to its remarkable local variability, making a correct interpretation of the data very difficult in the case of extremely small nanofeatures. Regular arrays of parallel 1D straight nanoprotrusions were fabricated by means of a low-dose focused ion beam, taking advantage of the amorphization-related swelling effect. The tribological properties of the patterns were detected in the presence of air and in vacuum (dry ambient) by atomic force microscopy. We have introduced a novel procedure and data analysis to reduce the uncertainties related to contact instabilities. The real time estimation of the radius of curvature of the contacting asperity enables us to study the dependence of the tribological properties of the patterns from their geometrical characteristics. The effect of the patterns on both adhesion and the coefficient of friction strongly depends on the contact area, which is linked to the local radius of curvature of the probe. However, a detectable hydrophobic character induced on the hydrophilic native SiO2 has been observed as well. The results suggest a scenario for capillary formation on the patterns.

  1. Study of longitudinal multibunch instabilities for LHC-type beams at the CERN Proton Synchrotron

    CERN Document Server

    Ventura, Letizia; Migliorati, Mauro; Palumbo, Luigi

    This Master thesis work has been carried out at CERN in the framework of the LHC (Large Hadron Collider) Injector upgrade program (LIU). Longitudinal coupled-bunch (CB) oscillations are an important limitation for the high-brightness beam accelerated in the CERN Proton Synchrotron. Up to present intensities they are suppressed by a dedicated feedback system limited to the first two dominant oscillation modes. In view of the proposed installation of a new wide-band FB system in the framework of the LIU program, measurements have been performed on the old system with the aim of dimensioning the new one. A new simulation program, called LCBC ( Longitudinal Coupled Bunch Simulation), has been used to study the behaviour of the CB FB. By means of this code I have started an extensive simulation campaign to benchmark the code with the theory of coupled bunch and to confirm that the 10 MHz cavity system is the main cause of the coupled bunch instabilities in the CERN PS.

  2. AFM-based tribological study of nanopatterned surfaces: the influence of contact area instabilities

    International Nuclear Information System (INIS)

    Rota, A; Serpini, E; Gazzadi, G C; Valeri, S

    2016-01-01

    Although the importance of morphology on the tribological properties of surfaces has long been proved, an exhaustive understanding of nanopatterning effects is still lacking due to the difficulty in both fabricating ‘really nano-’ structures and detecting their tribological properties. In the present work we show how the probe–surface contact area can be a critical parameter due to its remarkable local variability, making a correct interpretation of the data very difficult in the case of extremely small nanofeatures. Regular arrays of parallel 1D straight nanoprotrusions were fabricated by means of a low-dose focused ion beam, taking advantage of the amorphization-related swelling effect. The tribological properties of the patterns were detected in the presence of air and in vacuum (dry ambient) by atomic force microscopy. We have introduced a novel procedure and data analysis to reduce the uncertainties related to contact instabilities. The real time estimation of the radius of curvature of the contacting asperity enables us to study the dependence of the tribological properties of the patterns from their geometrical characteristics. The effect of the patterns on both adhesion and the coefficient of friction strongly depends on the contact area, which is linked to the local radius of curvature of the probe. However, a detectable hydrophobic character induced on the hydrophilic native SiO 2 has been observed as well. The results suggest a scenario for capillary formation on the patterns. (paper)

  3. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  4. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  5. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  6. CESR Conversion Damping Ring Studies of Electron Cloud Instabilities (CESR-TA)

    International Nuclear Information System (INIS)

    Rubin, David L.; Palmer, Mark A.

    2011-01-01

    In the International Linear Collider, two linear accelerators will accelerate bunches of positrons and electrons to over a hundred billion electron volts and collide them in a central detector. In order to obtain useful collision rates, the bunches, each containing twenty billion particles, must be compressed to a cross section of a few nanometers by a few hundred nanometers. In order to prepare these ultra high density bunches, damping rings (DRs) are employed before the linear accelerators. The DRs take the high emittance bunches that are provided by the electron and positron sources and, through the process of radiation damping, squeeze them into ultra low emittance beams that are ready for the main linear accelerators. In the damping rings, a number of effects can prevent the successful preparation of the beams. In the electron ring, an effect known as the fast ion instability can lead to beam growth and, in the positron ring, the build-up of an electron cloud (EC), which interacts with the circulating bunches, can produce the same effect. EC build-up and the subsequent interaction of the cloud with the positron beam in the DR have been identified as major risks for the successful construction of a linear collider. The CESRTA research program at the Cornell Electron Storage Ring (CESR) was developed in order to study the build-up of the EC, the details of its impact on ultra low emittance beams, as well as methods to mitigate the impact of the cloud. In the DR, the EC forms when synchrotron photons radiated from the circulating beam strike the walls of the vacuum chamber, resulting in the emission of photoelectrons. These low energy electrons can be accelerated across the vacuum chamber by the electric field of the beam, and strike the walls, causing the emission of secondary electrons. The secondary electrons are subsequently accelerated into the walls yet again via the same mechanism. The result is that the EC can rapidly begin to fill the vacuum chamber. In

  7. CESR Conversion Damping Ring Studies of Electron Cloud Instabilities (CESR-TA)

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, David L.; Palmer, Mark A.

    2011-08-02

    In the International Linear Collider, two linear accelerators will accelerate bunches of positrons and electrons to over a hundred billion electron volts and collide them in a central detector. In order to obtain useful collision rates, the bunches, each containing twenty billion particles, must be compressed to a cross section of a few nanometers by a few hundred nanometers. In order to prepare these ultra high density bunches, damping rings (DRs) are employed before the linear accelerators. The DRs take the high emittance bunches that are provided by the electron and positron sources and, through the process of radiation damping, squeeze them into ultra low emittance beams that are ready for the main linear accelerators. In the damping rings, a number of effects can prevent the successful preparation of the beams. In the electron ring, an effect known as the fast ion instability can lead to beam growth and, in the positron ring, the build-up of an electron cloud (EC), which interacts with the circulating bunches, can produce the same effect. EC build-up and the subsequent interaction of the cloud with the positron beam in the DR have been identified as major risks for the successful construction of a linear collider. The CESRTA research program at the Cornell Electron Storage Ring (CESR) was developed in order to study the build-up of the EC, the details of its impact on ultra low emittance beams, as well as methods to mitigate the impact of the cloud. In the DR, the EC forms when synchrotron photons radiated from the circulating beam strike the walls of the vacuum chamber, resulting in the emission of photoelectrons. These low energy electrons can be accelerated across the vacuum chamber by the electric field of the beam, and strike the walls, causing the emission of secondary electrons. The secondary electrons are subsequently accelerated into the walls yet again via the same mechanism. The result is that the EC can rapidly begin to fill the vacuum chamber. In

  8. Streaming gravity mode instability

    International Nuclear Information System (INIS)

    Wang Shui.

    1989-05-01

    In this paper, we study the stability of a current sheet with a sheared flow in a gravitational field which is perpendicular to the magnetic field and plasma flow. This mixing mode caused by a combined role of the sheared flow and gravity is named the streaming gravity mode instability. The conditions of this mode instability are discussed for an ideal four-layer model in the incompressible limit. (author). 5 refs

  9. A simulation study of electron-cloud instability and beam-induced multipacting in the LHC

    International Nuclear Information System (INIS)

    Zimmermann, F.

    1997-02-01

    In the LHC beam pipe, photoemission and secondary emission give rise to a quasi-stationary electron cloud, which is established after a few bunch passages. The response of this electron cloud to a transversely displaced bunch resembles a short-range wakefield and can cause a fast instability. In addition, beam-induced multipacting of the electrons may lead to an enhanced gas desorption and an associated pressure increase. In this paper the authors report preliminary simulation results of the electron-cloud build-up both in a dipole magnet and in a straight section of the LHC at top energy. The effective wakefield created by the electron cloud translates into an instability rise time of about 25 ms horizontally and 130 ms vertically. This rise time is not much larger than that of the resistive-wall instability at injection energy

  10. Housing Instability Is as Strong a Predictor of Poor Health Outcomes as Level of Danger in an Abusive Relationship: Findings from the SHARE Study

    Science.gov (United States)

    Rollins, Chiquita; Glass, Nancy E.; Perrin, Nancy A.; Billhardt, Kris A.; Clough, Amber; Barnes, Jamie; Hanson, Ginger C.; Bloom, Tina L.

    2012-01-01

    Advocates, clinicians, policy makers, and survivors frequently cite intimate partner violence (IPV) as an immediate cause of or precursor to housing problems. Research has indicated an association between homelessness and IPV, yet few studies examine IPV and housing instability. Housing instability differs from homelessness, in that someone…

  11. Sternal instability measured with radiostereometric analysis. A study of method feasibility, accuracy and precision

    DEFF Research Database (Denmark)

    Vestergaard, Rikke Falsig; Søballe, Kjeld; Hasenkam, John Michael

    2018-01-01

    BACKGROUND: A small, but unstable, saw-gap may hinder bone-bridging and induce development of painful sternal dehiscence. We propose the use of Radiostereometric Analysis (RSA) for evaluation of sternal instability and present a method validation. METHODS: Four bone analogs (phantoms) were sterno...... modality feasible for clinical evaluation of sternal stability in research. TRIAL REGISTRATION: ClinicalTrials.gov Identifier: NCT02738437 , retrospectively registered.......BACKGROUND: A small, but unstable, saw-gap may hinder bone-bridging and induce development of painful sternal dehiscence. We propose the use of Radiostereometric Analysis (RSA) for evaluation of sternal instability and present a method validation. METHODS: Four bone analogs (phantoms) were...

  12. Saha and temperature relaxation approximations for the study of ionization instability in partially ionized plasma

    International Nuclear Information System (INIS)

    Numano, M.

    1976-01-01

    The growth rates for the ionization instability obtained using the Saha and temperature relaxation approximations are compared with those obtained from an exact treatment, and the requirements for validity of these two approximations are obtained analytically. For the range of plasma parameters pertinent to MHD power generation it is found that the Saha approximation is valid for relatively high electron temperature, which it becomes inapplicable as the electron temperature is decreased. On the other hand, the temperature relaxation approximation is accurate over a wide range of electron temperature. It is found also that the marginal condition for the ionization instability is correctly obtained from both approximations. (author)

  13. Deviations in gait metrics in patients with chronic ankle instability: a case control study.

    Science.gov (United States)

    Gigi, Roy; Haim, Amir; Luger, Elchanan; Segal, Ganit; Melamed, Eyal; Beer, Yiftah; Nof, Matityahu; Nyska, Meir; Elbaz, Avi

    2015-01-01

    Gait metric alterations have been previously reported in patients suffering from chronic ankle instability (CAI). Previous studies of gait in this population have been comprised of relatively small cohorts, and the findings of these studies are not uniform. The objective of the present study was to examine spatiotemporal gait metrics in patients with CAI and examine the relationship between self-reported disease severity and the magnitude of gait abnormalities. Forty-four patients with CAI were identified and compared to 53 healthy controls. Patients were evaluated with spatiotemporal gait analysis via a computerized mat and with the Short Form (SF) - 36 health survey. Patients with CAI were found to walk with approximately 16% slower walking velocity, 9% lower cadence and approximately 7% lower step length. Furthermore, the base of support, during walking, in the CAI group was approximately 43% wider, and the single limb support phase was 3.5% shorter compared to the control group. All of the SF-36 8-subscales, as well as the SF-36 physical component summary and SF-36 mental component summary, were significantly lower in patients with CAI compared to the control group. Finally, significant correlations were found between most of the objective gait measures and the SF-36 mental component summary and SF-36 physical component summary. The results outline a gait profile for patients suffering from CAI. Significant differences were found in most spatiotemporal gait metrics. An important finding was a significantly wider base of support. It may be speculated that these gait alterations may reflect a strategy to deal with imbalance and pain. These findings suggest the usefulness of gait metrics, alongside with the use of self-evaluation questionnaires, in assessing disease severity of patients with CAI.

  14. Computational parametric study of a Richtmyer-Meshkov instability for an inclined interface.

    Science.gov (United States)

    McFarland, Jacob A; Greenough, Jeffrey A; Ranjan, Devesh

    2011-08-01

    A computational study of the Richtmyer-Meshkov instability for an inclined interface is presented. The study covers experiments to be performed in the Texas A&M University inclined shock tube facility. Incident shock wave Mach numbers from 1.2 to 2.5, inclination angles from 30° to 60°, and gas pair Atwood numbers of ∼0.67 and ∼0.95 are used in this parametric study containing 15 unique combinations of these parameters. Qualitative results are examined through a time series of density plots for multiple combinations of these parameters, and the qualitative effects of each of the parameters are discussed. Pressure, density, and vorticity fields are presented in animations available online to supplement the discussion of the qualitative results. These density plots show the evolution of two main regions in the flow field: a mixing region containing driver and test gas that is dominated by large vortical structures, and a more homogeneous region of unmixed fluid which can separate away from the mixing region in some cases. The interface mixing width is determined for various combinations of the parameters listed at the beginning of the Abstract. A scaling method for the mixing width is proposed using the interface geometry and wave velocities calculated using one-dimensional gas dynamic equations. This model uses the transmitted wave velocity for the characteristic velocity and an initial offset time based on the travel time of strong reflected waves. It is compared to an adapted Richtmyer impulsive model scaling and shown to scale the initial mixing width growth rate more effectively for fixed Atwood number.

  15. 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

  16. An experimental study of the effect of a pilot flame on technically pre-mixed, self-excited combustion instabilities

    Science.gov (United States)

    O'Meara, Bridget C.

    Combustion instabilities are a problem facing the gas turbine industry in the operation of lean, pre-mixed combustors. Secondary flames known as "pilot flames" are a common passive control strategy for eliminating combustion instabilities in industrial gas turbines, but the underlying mechanisms responsible for the pilot flame's stabilizing effect are not well understood. This dissertation presents an experimental study of a pilot flame in a single-nozzle, swirl-stabilized, variable length atmospheric combustion test facility and the effect of the pilot on combustion instabilities. A variable length combustor tuned the acoustics of the system to excite instabilities over a range of operating conditions without a pilot flame. The inlet velocity was varied from 25 -- 50 m/s and the equivalence ratio was varied from 0.525 -- 0.65. This range of operating conditions was determined by the operating range of the combustion test facility. Stability at each operating condition and combustor length was characterized by measurements of pressure oscillations in the combustor. The effect of the pilot flame on the magnitude and frequency of combustor stability was then investigated. The mechanisms responsible for the pilot flame effect were studied using chemiluminescence flame images of both stable and unstable flames. Stable flame structure was investigated using stable flame images of CH* chemiluminescence emission. The effect of the pilot on stable flame metrics such as flame length, flame angle, and flame width was investigated. In addition, a new flame metric, flame base distance, was defined to characterize the effect of the pilot flame on stable flame anchoring of the flame base to the centerbody. The effect of the pilot flame on flame base anchoring was investigated because the improved stability with a pilot flame is usually attributed to improved flame anchoring through the recirculation of hot products from the pilot to the main flame base. Chemiluminescence images

  17. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  18. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  19. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

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

    2010-01-01

    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). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  20. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  1. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  2. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    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)

  3. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  4. 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.

  5. 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)

  6. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    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

  7. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  8. Study of the internal structure, instabilities, and magnetic fields in the dense Z-pinch

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir V. [Univ. of Nevada, Reno, NV (United States)

    2016-08-17

    Z-pinches are sources of hot dense plasma which generates powerful x-ray bursts and can been applied to various areas of high-energy-density physics (HEDP). The 26-MA Z machine is at the forefront of many of these applications, but important aspects of HEDP have been studied on generators at the 1 MA current level. Recent development of laser diagnostics and upgrade of the Leopard laser at Nevada Terawatt Facility (NTF) give new opportunities for the dense Z-pinch study. The goal of this project is the investigation of the internal structure of the stagnated Z pinch including sub-mm and micron-scale instabilities, plasma dynamics, magnetic fields, and hot spots formation and initiation. New plasma diagnostics will be developed for this project. A 3D structure and instabilities of the pinch will be compared with 3D MHD and spectroscopic modeling and theoretical analysis. The structure and dynamics of stagnated Z pinches has been studied with x-ray self-radiation diagnostics which derive a temperature map of the pinch with a spatial resolution of 70-150 µm. The regular laser diagnostics at 532 nm does not penetrate in the dense pinch due to strong absorption and refraction in trailing plasma. Recent experiments at NTF showed that shadowgraphy at the UV wavelength of 266 nm unfolds a fine structure of the stagnated Z-pinch with unprecedented detail. We propose to develop laser UV diagnostics for Z pinches with a spatial resolution <5 μm to study the small-scale plasma structures, implement two-frame shadowgraphy/interferometry, and develop methods for investigation of strong magnetic fields. New diagnostics will help to understand better basic physical processes in Z pinches. A 3D internal structure of the pinch and characteristic instabilities will be studied in wire arrays with different configurations and compared with 3D MHD simulations and analytical models. Mechanisms of “enhanced heating” of Z-pinch plasma will be studied. Fast dynamics of stagnated

  9. A comparative study of the single-mode Richtmyer-Meshkov instability

    Science.gov (United States)

    Bai, X.; Deng, X.-L.; Jiang, L.

    2017-11-01

    In this work, the single-mode Richtmyer-Meshkov instability is studied numerically to find a reasonable nonlinear theoretical model which can be applied to predict the interface evolution from the linear stage to the early nonlinear stage. The cut-cell-based sharp-interface methods MuSiC+ (Chang et al. in J Comput Phys 242:946-990, 2013) and CCGF (Bai and Deng in Adv Appl Math Mech 9(5):1052-1075, 2017) are applied to generate numerical results for comparisons. Classical Air-SF6 and Air-Helium conditions are applied in this study, and initial amplitude and Atwood number are varied for comparison. Comparisons to the simulation results from the literature show the applicability of MuSiC+ and CCGF. Comparisons to the nonlinear theoretical models show that ZS (Zhang and Sohn in Phys Lett A 212:149-155, 1996; Phys Fluids 9:1106-1124, 1997), SEA (Sadot et al. in Phys Rev Lett 80:1654-1657, 1998), and DR (Dimonte and Ramaprabhu in Phys Fluids 22:014104, 2010) models are valid for both spike and bubble growth rates, and MIK (Mikaelian in Phys Rev E 67:026319, 2003) and ZG (Zhang and Guo in J Fluid Mech 786:47-61, 2016) models are valid for bubble growth rate, when the initial perturbation is small and the Atwood number is low, but only the DR model is applicable for both spike and bubble growth rates when the initial perturbation amplitude and the Atwood number are large. A new term of non-dimensional initial perturbation amplitude is presented and multiplied to the DR model to get a unified fitted DR model, which gives consistent results to the simulation ones for small and large initial amplitudes.

  10. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-01-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  11. Study of the internal structure, instabilities, and magnetic fields in the dense Z-pinch

    International Nuclear Information System (INIS)

    Ivanov, Vladimir V.

    2016-01-01

    Z-pinches are sources of hot dense plasma which generates powerful x-ray bursts and can been applied to various areas of high-energy-density physics (HEDP). The 26-MA Z machine is at the forefront of many of these applications, but important aspects of HEDP have been studied on generators at the 1 MA current level. Recent development of laser diagnostics and upgrade of the Leopard laser at Nevada Terawatt Facility (NTF) give new opportunities for the dense Z-pinch study. The goal of this project is the investigation of the internal structure of the stagnated Z pinch including sub-mm and micron-scale instabilities, plasma dynamics, magnetic fields, and hot spots formation and initiation. New plasma diagnostics will be developed for this project. A 3D structure and instabilities of the pinch will be compared with 3D MHD and spectroscopic modeling and theoretical analysis. The structure and dynamics of stagnated Z pinches has been studied with x-ray self-radiation diagnostics which derive a temperature map of the pinch with a spatial resolution of 70-150 µm. The regular laser diagnostics at 532 nm does not penetrate in the dense pinch due to strong absorption and refraction in trailing plasma. Recent experiments at NTF showed that shadowgraphy at the UV wavelength of 266 nm unfolds a fine structure of the stagnated Z-pinch with unprecedented detail. We propose to develop laser UV diagnostics for Z pinches with a spatial resolution 20 MG, suggested in micropinches, Cotton-Mouton and cutoff diagnostics will be applied. A picosecond optical Kerr shutter will be tested to increase a sensitivity of UV methods for application at multi-MA Z pinches. The proposal is based on the experimental capability of NTF. The Zebra generator produces 1-1.7 MA Z-pinches with electron plasma density of 10"2"0-10"2"1cm"-"3, electron temperature of 0.5-1 keV, and magnetic fields >10 MG. The Leopard laser was upgraded to energy of 90-J at 0.8 ns. This regime will be used for laser initiation

  12. The Experimental Study of Rayleigh-Taylor Instability using a Linear Induction Motor Accelerator

    Science.gov (United States)

    Yamashita, Nicholas; Jacobs, Jeffrey

    2009-11-01

    The experiments to be presented utilize an incompressible system of two stratified miscible liquids of different densities that are accelerated in order to produce the Rayleigh-Taylor instability. Three liquid combinations are used: isopropyl alcohol with water, a calcium nitrate solution or a lithium polytungstate solution, giving Atwood numbers of 0.11, 0.22 and 0.57, respectively. The acceleration required to drive the instability is produced by two high-speed linear induction motors mounted to an 8 m tall drop tower. The motors are mounted in parallel and have an effective acceleration length of 1.7 m and are each capable of producing 15 kN of thrust. The liquid system is contained within a square acrylic tank with inside dimensions 76 x76x184 mm. The tank is mounted to an aluminum plate, which is driven by the motors to create constant accelerations in the range of 1-20 g's, though the potential exists for higher accelerations. Also attached to the plate are a high-speed camera and an LED backlight to provide continuous video of the instability. In addition, an accelerometer is used to provide acceleration measurements during each experiment. Experimental image sequences will be presented which show the development of a random three-dimensional instability from an unforced initial perturbation. Measurements of the mixing zone width will be compared with traditional growth models.

  13. Microbunching Instability Effect Studies and Laser Heater Optimization for the SPARX FEL Accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Vaccarezza, C.; Chiadroni, E.; Ferrario, M.; Giannessi, L.; Quattromini, M.; Ronsivalle, C.; Venturini, C.; Migliorati, M.; Dattoli, G.

    2010-05-23

    The effects of microbunching instability for the SPARX accelerator have been analyzed by means of numerical simulations. The laser heater counteracting action has been addressed in order to optimize the parameters of the compression system, either hybrid RF plus magnetic chicane or only magnetic, and possibly enhance the FEL performance.

  14. Study on the relationship between DNA-PKcs and genomic instability and hyper-radiosensitivity

    International Nuclear Information System (INIS)

    Yang Kang; Zhu Jiayun; Ding Nan; Li Junhong; Hu Wentao; Su Fengtao; He Jinpeng; Li Sha

    2010-01-01

    To investigate the relationship between DNA-PKcs and genome instability and hyper-radiosensitivity, human glioma cell lines M059K and M059J, as a model expressing wild-type DNA-PKcs and a model defective in DNA-PKcs activity, were exposed to low doses of X-rays. Cells survival fractions were assessed by colony-forming assay and Cytochalasin-B micronucleus assay was employed to detect the genomic instability happening in each single irradiated colony. It has been found that as the post-incubation time increased, M059K cells expressing wild-type DNA-PKcs exhibited low-dose hyper-radiosensitivity and showed a similar genomic instability after 0.2 Gy and 0.6 Gy irradiations, but the M059J cells lacking in DNA-PKcs didn't present low-dose hyper-radiosensitivity and showed a higher genomic instability of 0.6 Gy than that of 0.2 Gy. The results indicate that DNA-PKcs may act as one of the key factors that lead to low-dose hyper-radiosensitivity. (authors)

  15. Core concept for long operating cycle simplified BWR (LSBWR)

    International Nuclear Information System (INIS)

    Kouji, Hiraiwa; Noriyuki, Yoshida; Mikihide, Nakamaru; Hideaki, Heki; Masanori, Aritomi

    2002-01-01

    An innovative core concept for a long operating cycle simplified BWR (LSBWR) is currently being developed under a Toshiba Corporation and Tokyo Institute of Technology joint study. In this core concept, the combination of enriched uranium oxide fuels and loose-pitched lattice is adopted for an easy application of natural circulation. A combination of enriched gadolinium and 0.7-times sized small bundle with peripheral-positioned gadolinium rod is also adopted as a key design concept for 15-year cycle operation. Based on three-dimensional nuclear and thermal hydraulic calculation, a nuclear design for fuel bundle has been determined. Core performance has been evaluated based on this bundle design and shows that thermal performance and reactivity characteristics meet core design criteria. Additionally, a control rod operation plan for an extension of control rod life has been successfully determined. (author)

  16. Cobra-TF simulation of BWR bundle dry out experiments

    Energy Technology Data Exchange (ETDEWEB)

    Frepoli, C.; Ireland, A.; Hochreiter, L.; Ivanov, K. [Penn State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, PA (United States); Velten, R. [Siemens Nuclear Power GmbH, Erlangen (Germany)

    2001-07-01

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

  17. Availability analysis of United States BWR IV electrical generation plants

    International Nuclear Information System (INIS)

    Renick, D.H.; Li, F.; Todreas, N.E.

    1998-01-01

    Availability, as quantified by power output levels, from all active U.S. BWR IV plants were analyzed over a seven and a half year period to determine the operational characteristics of these plants throughout an operating cycle. The operational data were examined for infant mortality, end of cycle decreased availability, and seasonal availability variations. Scheduled outages were also examined to determine the industry's current approach to planning maintenance outages. The results of this study show that nuclear power plants do suffer significant infant mortality following a refueling outage. And while they do not suffer an end of cycle decrease in availability, a mid-cycle period of decreased availability is evident. This period of decreased availability is due to a combination of increased forced unavailability and seasonally scheduled maintenance and refueling outages. These findings form the start of a rational approach to increasing plant availability. (author)

  18. 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)

  19. Studies of the charge instabilities in the complex nano-objects: clusters and bio-molecular systems

    International Nuclear Information System (INIS)

    Manil, B.

    2007-11-01

    For the last 6 years, my main research works focused on i) the Coulomb instabilities and the fragmentation processes of fullerenes and clusters of fullerenes ii) the stability and the reactivity of complex bio-molecular systems. Concerning the clusters of fullerenes, which are van der Waals type clusters, we have shown that the multiply charged species, obtained in collisions with slow highly charged ions, keep their structural properties but become very good electric conductor. In another hand, with the aim to understand the role of the biologic environment at the molecular scale in the irradiation damage of complex biomolecules, we have studied the charge stabilities of clusters of small biomolecules and the dissociation processes of larger nano-hydrated biomolecules. Theses studies have shown that first, specific molecular recognition mechanisms continue to exist in gas phase and secondly, a small and very simple biochemical environment is enough to change the dynamics of instabilities. (author)

  20. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    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 {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs 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.

  1. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  2. Clinical study of bilateral decompression via vertebral lamina fenestration for lumbar interbody fusion in the treatment of lower lumbar instability

    OpenAIRE

    GUO, SHUGUANG; SUN, JUNYING; TANG, GENLIN

    2013-01-01

    The aim of this study was to observe the clinical effects of bilateral decompression via vertebral lamina fenestration for lumbar interbody fusion in the treatment of lower lumbar instability. The 48 patients comprised 27 males and 21 females, aged 47?72 years. Three cases had first and second degree lumbar spondylolisthesis and all received bilateral vertebral lamina fenestration for posterior lumbar interbody fusion (PLIF) using a threaded fusion cage (TFC), which maintains the three-column...

  3. A 10 year follow-up study after Roux-Elmslie-Trillat treatment for cases of patellar instability

    OpenAIRE

    Endres, Stefan; Wilke, Axel

    2011-01-01

    Abstract Background A retrospective study concerning patients presenting with patella instability, treated using a Roux-Elmslie-Trillat reconstruction operation and followed up for 10 years following surgery, is presented. Methods Pre-operative and follow-up radiographic evaluation included the weight-bearing anteroposterior and merchant views. Evaluation was carried out using the Insall-Salvati index, sulcus and congruence angle. The Roux-Elmslie-Trillat reconstruction operation was performe...

  4. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Science.gov (United States)

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

  5. TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kozlowski, T. [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, T.; Xu, Y.; Wysocki, A. [Univ. of Michigan, Ann Arbor, MI (United States); Ivanov, K.; Magedanz, J.; Hardgrove, M. [Pennsylvania State Univ., Univ. Park, PA (United States); March-Leuba, J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hudson, N.; Woodyatt, D. [Nuclear Regulatory Commission, Rockville, MD (United States)

    2012-07-01

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)

  6. TRACE/PARCS validation for BWR stability based on OECD/NEA Oskarshamn-2 benchmark

    International Nuclear Information System (INIS)

    Kozlowski, T.; Roshan, S.; Lefvert, T.; Downar, T.; Xu, Y.; Wysocki, A.; Ivanov, K.; Magedanz, J.; Hardgrove, M.; Netterbrant, C.; March-Leuba, J.; Hudson, N.; Sandervag, O.; Bergman, A.

    2011-01-01

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event, which culminated in diverging power oscillations with decay ratio greater than 1.3. The event was successfully modeled by TRACE/PARCS coupled code system and the details of the modeling and solution are described in the paper. The obtained results show excellent agreement with the plant data, capturing the entire behavior of the transient including onset of instability, growth of oscillation (decay ratio) and the oscillation frequency. The event allows coupled code validation for BWR with a real, challenging stability event, which challenges accuracy of neutron kinetics (NK), thermal-hydraulics (TH) and TH/NK coupling. The success of this work has demonstrated the ability of 3-D coupled code systems to capture the complex behavior of BWR stability events. The problem is released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (author)

  7. Thermal-hydraulics stability of natural circulation BWR under startup. Flashing effects

    International Nuclear Information System (INIS)

    Hu, Rui; Kazimi, Mujid S.

    2009-01-01

    To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a boiling natural circulation reactor like the ESBWR. In such systems, thermal-hydraulic stability during low pressure start-up should be examined while considering the flashing induced by the pressure drop in the channel and the chimney due to gravity head. In this work, a BWR stability analysis code in the frequency domain, named FISTAB (Flashing-Induced STability Analysis for BWR), was developed to address the issue of flashing-induced instability. A thermal-hydraulics non-homogeneous equilibrium model (NHEM) based on a drift flux formulation along with a lumped fuel dynamics model is incorporated in the work. The vapor generation rate is derived from the mixture energy conservation equation while considering the effect of flashing. The functionality of the FISTAB code was confirmed by comparison to experimental results from SIRIUS-N facility at CRIEPI, Japan. Both stationary and perturbation results agree well with the experimental results. (author)

  8. The possibility and the effects of a steam explosion in the BWR lower head on recriticality of a BWR core

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.

    2002-12-01

    The report describes an analysis considering a BWR postulated severe accident scenario during which the late vessel automatic depressurization brings the water below the level of the bottom core plate. The subsequent lack of ECCS leads to core heat up during which the control rods melt and the melt deposits on the core plate. At that point of time in the scenario, the core fuel bundles are still intact and the Zircaloy clad oxidation is about to start. The objective of the study is to provide the conditions of reflood into the hot core due to the level swell or a slug delivered from the lower head as the control rod melt drops into the water. These conditions are employed in the neutronic analysis with the RECRIT code to determine if the core recriticality may be achieved. (au)

  9. Managing BWR plant life extension

    International Nuclear Information System (INIS)

    Ianni, P.W.; Kiss, E.

    1985-01-01

    Recent studies have confirmed that extending the useful life of a large nuclear plant can be justified with very high cost benefit ratio. In turn, experience with large power plant systems and equipment has shown that a well-integrated and -managed plan is essential in order to achieve potential economic benefits. Consequently, General Electric's efforts have been directed at establishing a life extension plan that considers alternative options and cost-effective steps that can be taken in early life, those appropriate during middle life, and those required in late life. This paper briefly describes an approach designed to provide the plant owner a maximum of flexibility in developing a life extension plan

  10. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  11. Simulation study of electron cloud induced instabilities and emittance growth for the CERN Large Hadron Collider proton beam

    CERN Document Server

    Benedetto, Elena; Schulte, Daniel; Rumolo, Giovanni

    2005-01-01

    The electron cloud may cause transverse single-bunch instabilities of proton beams such as those in the Large Hadron Collider (LHC) and the CERN Super Proton Synchrotron (SPS). We simulate these instabilities and the consequent emittance growth with the code HEADTAIL, which models the turn-by-turn interaction between the cloud and the beam. Recently some new features were added to the code, in particular, electric conducting boundary conditions at the chamber wall, transverse feedback, and variable beta functions. The sensitivity to several numerical parameters has been studied by varying the number of interaction points between the bunch and the cloud, the phase advance between them, and the number of macroparticles used to represent the protons and the electrons. We present simulation results for both LHC at injection and SPS with LHC-type beam, for different electron-cloud density levels, chromaticities, and bunch intensities. Two regimes with qualitatively different emittance growth are observed: above th...

  12. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    Shack, W.J.; Kassner, T.F.; Maiya, P.S.; Park, J.Y.; Ruther, W.; Kuzay, T.; Rybicki, E.F.; Stonesifer, R.B.

    1988-01-01

    Piping in light-water-reactor power systems has been affected by several types of environmental degradation. This paper presents results from studies of (1) stress corrosion crack growth in fracture mechanics specimens of modified Type 347 SS and Type 304/308L SS weld overlay material, (2) heat-to-heat variations in stress corrosion cracking (SCC) of Types 316NG and 347 SS, (3) SCC of sensitized Type 304 SS in water with cupric ion or organic acid impurities, (4) electrochemical potential (ECP) measurements under gamma irradiation, (5) SCC of ferritic steels, (6) strain-controlled fatigue of Type 316NG SS in air at ambient temperature, and (7) through-wall residual stress measurements and finite-element calculation of residual stresses in weldments treated by a mechanical stress improvement process (MSIP). Fracture-mechanics crack-growth-rate tests on Type 316NG SS have shown that transgranular cracking can occur even in high purity environments, whereas no crack growth was observed in Type 347 SS even in impurity environments. In tests on weld overlay specimens, no cracks penetrated into the overlay even in impurity environments. Instead, the cracks branched when they approached the overlay, and then grew parallel to interface. In SCC tests on sensitized Type 304 SS, cupric ions at concentrations greater than ∼1 ppm were found to be deleterious, whereas organic acids at this concentration were not detrimental. Tests on several ferritic steels indicate a strong correlation between the sulfur content of the steels and susceptibility to SCC. External gamma radiation fields produced a large positive shift in the ECP of Type 304 SS at low dissolved-oxygen concentrations (<5 ppb), whereas in the absence of an external gamma field there was no difference in the ECP values of irradiated and nonirradiated material. Fatigue data for Type 316NG SS are consistent with the ASME code mean curve at high strains, but fall below the curve at low strains. Calculations of the

  13. Resonant Drag Instabilities in protoplanetary disks: the streaming instability and new, faster-growing instabilities

    Science.gov (United States)

    Squire, Jonathan; Hopkins, Philip F.

    2018-04-01

    We identify and study a number of new, rapidly growing instabilities of dust grains in protoplanetary disks, which may be important for planetesimal formation. The study is based on the recognition that dust-gas mixtures are generically unstable to a Resonant Drag Instability (RDI), whenever the gas, absent dust, supports undamped linear modes. We show that the "streaming instability" is an RDI associated with epicyclic oscillations; this provides simple interpretations for its mechanisms and accurate analytic expressions for its growth rates and fastest-growing wavelengths. We extend this analysis to more general dust streaming motions and other waves, including buoyancy and magnetohydrodynamic oscillations, finding various new instabilities. Most importantly, we identify the disk "settling instability," which occurs as dust settles vertically into the midplane of a rotating disk. For small grains, this instability grows many orders of magnitude faster than the standard streaming instability, with a growth rate that is independent of grain size. Growth timescales for realistic dust-to-gas ratios are comparable to the disk orbital period, and the characteristic wavelengths are more than an order of magnitude larger than the streaming instability (allowing the instability to concentrate larger masses). This suggests that in the process of settling, dust will band into rings then filaments or clumps, potentially seeding dust traps, high-metallicity regions that in turn seed the streaming instability, or even overdensities that coagulate or directly collapse to planetesimals.

  14. Experimental study of external kink instabilities in the Columbia High Beta Tokamak

    International Nuclear Information System (INIS)

    Ivers, T.H.

    1991-01-01

    The generation of power through controlled thermonuclear fusion reactions in a magnetically confined plasma holds promise as a means of supplying mankind's future energy needs. The device most technologically advanced in pursuit of this goal is the tokamak, a machine in which a current-carrying toroidal plasma is thermally isolated from its surroundings by a strong magnetic field. To be viable, the tokamak reactor must produce a sufficiently large amount of power relative to that needed to sustain the fusion reactions. Plasma instabilities may severely limit this possibility. In this work, I describe experimental measurements of the magnetic structure of large-scale, rapidly-growing instabilities that occur in a tokamak when the current or pressure of the plasma exceeds a critical value relative to the magnetic field, and I compare these measurements with theoretical predictions

  15. Theoretical studies of plasma turbulence and hydrodynamic instability in fusion targets. Summary report

    International Nuclear Information System (INIS)

    Morse, R.L.

    1984-01-01

    During the last year, the principle activities and accomplishments under this contract have been: (1) detailed examination of steady flow model calculations of ablation driven Taylor instability to determine the specific cause of growth rate reductions at short wavelengths; (2) implementation of the ''piggy back'' type, time-dependent, spherical harmonic, stability code, PANSY, on the Livermore Octopus system, and expansion of the code to treat a wider range of problems; and (3) use of the PANSY stability code to confirm, with this more realistic model, the results of Item 1 above and to begin learning about the phenomenology of Taylor mode growth and propagation, and the phenomenology of the development of the Richtmyer-Meshkov shock instability in compressible shells

  16. Analysis of radiological consequences in a typical BWR with a mark-II containment

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro

    2003-01-01

    INS/NUPEC in Japan has been carrying out the Level 3 PSA program. In the program, the MACCS2 code has been extensively applied to analyze radiological consequences for typical BWR and PWR plants in Japan. The present study deals with analysis of effects of the AMs, which were implemented by industries, on radiological consequence for a typical BWR with a Mark-II containment. In the present study, source terms and their frequencies of source terms were used based on results of Level 2 PSA taking into account AM countermeasures. Radiological consequences were presented with dose risks (Sv/ry), which were multiplied doses (Sv) by containment damage frequencies (/ry), and timing of radionuclides release to the environment. The results of the present study indicated that the dose risks became negligible in most cases taking AM countermeasures and evacuations. (author)

  17. A Study of Current Driven Electrostatic Instability on the Auroral Zone

    Directory of Open Access Journals (Sweden)

    S. Y. Kim

    1986-12-01

    Full Text Available According to recent satellite observations, strong ion transverse acceleration to the magnetic field(ion conics has been known. The ion conics may be a result of electrostatic waves frequently observed on the auroral zone. Both linear and nonlinear theory of electrostatic instability driven by an electron current based on 1-dimenstional particle simulation experiment have been considered. From the results of simulation strong ion transverse acceleration has been shown.

  18. Structural instabilities in superconductor Y-Ba-Cu-O studied by positron annihilation

    International Nuclear Information System (INIS)

    Wang Shaojie; Li Xiaohua; Chen Yilong; Li Shiqing; Feng Guohua; Wang Zhu; Chen Ang; Li Biaorong

    1989-01-01

    The positron lifetime and Doppler broadening were measured as a function of temperature between 100K and 300K for Y-Ba-Cu-O superconductor. There are anomalous changes in the positron lifetimes and S parameters near 130K and 260K which imply that some phase transitions related to superconductivity happenned there. The results were discussed in terms of the lattice instabilities which may caused by the ordering readjustment of oxygen vacancies

  19. Fracture assessment of a main reactor coolant pump in a BWR with encountered defects

    Energy Technology Data Exchange (ETDEWEB)

    Brickstad, B [Swedish Plant Inspectorate, Stockholm (Sweden)

    1988-12-31

    This document presents a case-study fracture assessment in BWR type reactor components. A cast stainless steel presenting defects due to thermal is studied. The stress analysis performed by aid of a finite element technique shows that a Leak Before Break situation could be expected. Nevertheless, it may be concluded that the cross section of the pump where the defect area was located can withstand very deep cracks before the risk of failure becomes unacceptable. (TEC).

  20. Dose rate reduction method for NMCA applied BWR plants

    International Nuclear Information System (INIS)

    Nagase, Makoto; Aizawa, Motohiro; Ito, Tsuyoshi; Hosokawa, Hideyuki; Varela, Juan; Caine, Thomas

    2012-09-01

    BRAC (BWR Radiation Assessment and Control) dose rate is used as an indicator of the incorporation of activated corrosion by products into BWR recirculation piping, which is known to be a significant contributor to dose rate received by workers during refueling outages. In order to reduce radiation exposure of the workers during the outage, it is desirable to keep BRAC dose rates as low as possible. After HWC was adopted to reduce IGSCC, a BRAC dose rate increase was observed in many plants. As a countermeasure to these rapid dose rate increases under HWC conditions, Zn injection was widely adopted in United States and Europe resulting in a reduction of BRAC dose rates. However, BRAC dose rates in several plants remain high, prompting the industry to continue to investigate methods to achieve further reductions. In recent years a large portion of the BWR fleet has adopted NMCA (NobleChem TM ) to enhance the hydrogen injection effect to suppress SCC. After NMCA, especially OLNC (On-Line NobleChem TM ), BRAC dose rates were observed to decrease. In some OLNC applied BWR plants this reduction was observed year after year to reach a new reduced equilibrium level. This dose rate reduction trends suggest the potential dose reduction might be obtained by the combination of Pt and Zn injection. So, laboratory experiments and in-plant tests were carried out to evaluate the effect of Pt and Zn on Co-60 deposition behaviour. Firstly, laboratory experiments were conducted to study the effect of noble metal deposition on Co deposition on stainless steel surfaces. Polished type 316 stainless steel coupons were prepared and some of them were OLNC treated in the test loop before the Co deposition test. Water chemistry conditions to simulate HWC were as follows: Dissolved oxygen, hydrogen and hydrogen peroxide were below 5 ppb, 100 ppb and 0 ppb (no addition), respectively. Zn was injected to target a concentration of 5 ppb. The test was conducted up to 1500 hours at 553 K. Test

  1. Association of postural instability with asymptomatic cerebrovascular damage and cognitive decline: the Japan Shimanami health promoting program study.

    Science.gov (United States)

    Tabara, Yasuharu; Okada, Yoko; Ohara, Maya; Uetani, Eri; Kido, Tomoko; Ochi, Namiko; Nagai, Tokihisa; Igase, Michiya; Miki, Tetsuro; Matsuda, Fumihiko; Kohara, Katsuhiko

    2015-01-01

    Asymptomatic cerebral small-vessel disease (cSVD) in elderly individuals are potent risk factors for stroke. In addition to common clinical risk factors, postural instability has been postulated to be associated with cSVD in older frail patients. Here, we conducted a cross-sectional study to understand the possible link between postural instability and asymptomatic cSVD further, namely periventricular hyperintensity, lacunar infarction, and microbleeds, as well as cognitive function, in a middle-aged to elderly general population (n=1387). Postural instability was assessed based on one-leg standing time (OLST) and posturography findings. cSVD was evaluated by brain MRI. Mild cognitive impairment was assessed using a computer-based questionnaire, and carotid intima-media thickness as an index of atherosclerosis was measured via ultrasonography. Frequency of short OLST, in particular 2, 34.5%; microbleeds lesion: none, 10.1%; 1, 15.3%; >2, 30.0%; periventricular hyperintensity grade: 0, 5.7%; 1, 11.5%; >2, 23.7%). The association of short OLST with lacunar infarction and microbleeds but not periventricular hyperintensity remained significant even after adjustment for possible covariates (lacunar infarction, P=0.009; microbleeds, P=0.003; periventricular hyperintensity, P=0.601). In contrast, no significant association was found between posturographic parameters and cSVD, whereas these parameters were linearly associated with OLST. Short OLST was also significantly associated with reduced cognitive function independent of covariates, including cSVD (P=0.002). Postural instability was found to be associated with early pathological changes in the brain and functional decline, even in apparently healthy subjects. © 2014 American Heart Association, Inc.

  2. Best-estimate analysis development for BWR systems

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

    1986-01-01

    The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

  3. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

  4. BWR stability: history and state-of-the-art

    International Nuclear Information System (INIS)

    Yadigaroglu, George

    2014-01-01

    The paper briefly recalls the historical developments, reviews the important phenomena, the analytical and simulation tools that are used for the analysis of BWR stability focussing on the linear, frequency domain methods

  5. Thermal Shrinkage for Shoulder Instability

    OpenAIRE

    Toth, Alison P.; Warren, Russell F.; Petrigliano, Frank A.; Doward, David A.; Cordasco, Frank A.; Altchek, David W.; O’Brien, Stephen J.

    2010-01-01

    Thermal capsular shrinkage was popular for the treatment of shoulder instability, despite a paucity of outcomes data in the literature defining the indications for this procedure or supporting its long-term efficacy. The purpose of this study was to perform a clinical evaluation of radiofrequency thermal capsular shrinkage for the treatment of shoulder instability, with a minimum 2-year follow-up. From 1999 to 2001, 101 consecutive patients with mild to moderate shoulder instability underwent...

  6. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  7. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  8. Development of long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-01-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  9. BWR plant advanced central control panel PODIA

    International Nuclear Information System (INIS)

    Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

    1983-01-01

    BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

  10. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  11. A BWR Safety and Operability Improvements

    International Nuclear Information System (INIS)

    Sawyer, Craig D.

    1993-01-01

    The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

  12. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

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

    1990-01-01

    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)

  13. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

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

    1990-01-01

    BWR stability can be treated with reduced order 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 struct-ure 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 in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

  14. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  15. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Kato, Shigeru

    1996-01-01

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  16. Facility of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsuji

    1998-01-01

    A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

  17. Early Life Psychosocial Stressors and Housing Instability among Young Sexual Minority Men: the P18 Cohort Study.

    Science.gov (United States)

    Krause, Kristen D; Kapadia, Farzana; Ompad, Danielle C; D'Avanzo, Paul A; Duncan, Dustin T; Halkitis, Perry N

    2016-06-01

    Homelessness and housing instability is a significant public health problem among young sexual minority men. While there is a growing body of literature on correlates of homelessness among sexual minority men, there is a lack of literature parsing the different facets of housing instability. The present study examines factors associated with both living and sleeping in unstable housing among n = 600 sexual minority men (ages 18-19). Multivariate models were constructed to examine the extent to which sociodemographic, interpersonal, and behavioral factors as well as adverse childhood experiences explain housing instability. Overall, 13 % of participants reported sleeping in unstable housing and 18 % had lived in unstable housing at some point in the 6 months preceding the assessment. The odds of currently sleeping in unstable housing were greater among those who experienced more frequent lack of basic needs (food, proper hygiene, clothing) during their childhoods. More frequent experiences of childhood physical abuse and a history of arrest were associated with currently living in unstable housing. Current enrollment in school was a protective factor with both living and sleeping in unstable housing. These findings indicate that being unstably housed can be rooted in early life experiences and suggest a point of intervention that may prevent unstable housing among sexual minority men.

  18. Experimental study of linear and nonlinear regimes of density-driven instabilities induced by CO2 dissolution in water

    International Nuclear Information System (INIS)

    Outeda, R.; D'Onofrio, A.; El Hasi, C.; Zalts, A.

    2014-01-01

    Density driven instabilities produced by CO 2 (gas) dissolution in water containing a color indicator were studied in a Hele Shaw cell. The images were analyzed and instability patterns were characterized by mixing zone temporal evolution, dispersion curves, and the growth rate for different CO 2 pressures and different color indicator concentrations. The results obtained from an exhaustive analysis of experimental data show that this system has a different behaviour in the linear regime of the instabilities (when the growth rate has a linear dependence with time), from the nonlinear regime at longer times. At short times using a color indicator to see the evolution of the pattern, the images show that the effects of both the color indicator and CO 2 pressure are of the same order of magnitude: The growth rates are similar and the wave numbers are in the same range (0–30 cm −1 ) when the system is unstable. Although in the linear regime the dynamics is affected similarly by the presence of the indicator and CO 2 pressure, in the nonlinear regime, the influence of the latter is clearly more pronounced than the effects of the color indicator

  19. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices III and IV. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    The items listed below summarize the detail sections which follow: a listing of definitions and a discussion of the general treatment of data within the random variable approach as utilized by the study; a tabulation of the assessed data base containing failure classifications, final assessed ranges utilized in quantification and reference source values considered in determining the ranges; a discussion of nuclear power plant experience that was used to validate the data assessment by testing its applicability as well as to check on the adequacy of the model to incorporate typical real incidents; an expanded presentation of the data assessment giving information on applicability considerations; a discussion of test and maintenance data including comparisons of models with experience data; and special topics, including assessments required for the initiating event probabilities and human error data and modeling.

  20. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  1. BWR startup and shutdown activity transport control

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

    2010-07-01

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

  2. Pain, instability, and familial discord: a qualitative study into women who use drugs in Malaysia.

    Science.gov (United States)

    Rahman, Fifa; Lall, Priya; Iqbal, Sarah; Vicknasingam, B

    2015-11-05

    Out of 20,887 persons who use drugs that came into contact with the National Anti-Drugs Agency (NADA) officials in the year 2013, 3.2% were women. Because women who use drugs (WWUD) are often a hidden population, this may be an underestimate. International literature shows that women who use drugs face increased risk of HIV, intimate partner violence, and mental health issues. Similar literature in Malaysia is lacking, and thus, the objective of our study was exploratory in nature. Thirty-eight women who use drugs were interviewed using a semi-structured topic guide in Kelantan, Penang, Johor, Kuala Lumpur, and Selangor. Locations were chosen purposively. Nineteen women were interviewed individually and the remaining 19 were in focus group discussions (FGDs). All interviews were transcribed verbatim, translated to English, and analyzed with NVivo. Median age of respondents was 35.5 years old, 89.5% ethnic Malays, majority having married below the age of 20, and were of low socioeconomic backgrounds. Youngest age of initiation into drug use was 9 years old. Most reported is inhalation of amphetamine-type substances. Seven reported ever injecting. Three themes emerged: (a) repeating patterns of fluid family structures and instability; (b) "pain" and "difficulty" as features of home life; and (c) seeking marriage as a source of stabilization and practices of power within those marriages. Respondents often came from very fluid family environments and married to find stability, only to be drawn into a similar cycle. None of the women who had been separated from their children either institutionally, by family members, or by third parties, had accessed legal recourse for the loss of their parental rights. Unstable familial relationships or environments contributed to earlier initiation of drug use which raised questions about support services for WWUD and children who use drugs. Respondents were drawn into unstable and/or abusive relationships, perpetuating social

  3. Instabilities in mimetic matter perturbations

    Energy Technology Data Exchange (ETDEWEB)

    Firouzjahi, Hassan; Gorji, Mohammad Ali [School of Astronomy, Institute for Research in Fundamental Sciences (IPM), P.O. Box 19395-5531, Tehran (Iran, Islamic Republic of); Mansoori, Seyed Ali Hosseini, E-mail: firouz@ipm.ir, E-mail: gorji@ipm.ir, E-mail: shosseini@shahroodut.ac.ir, E-mail: shossein@ipm.ir [Physics Department, Shahrood University of Technology, P.O. Box 3619995161 Shahrood (Iran, Islamic Republic of)

    2017-07-01

    We study cosmological perturbations in mimetic matter scenario with a general higher derivative function. We calculate the quadratic action and show that both the kinetic term and the gradient term have the wrong sings. We perform the analysis in both comoving and Newtonian gauges and confirm that the Hamiltonians and the associated instabilities are consistent with each other in both gauges. The existence of instabilities is independent of the specific form of higher derivative function which generates gradients for mimetic field perturbations. It is verified that the ghost instability in mimetic perturbations is not associated with the higher derivative instabilities such as the Ostrogradsky ghost.

  4. Analysis of natural circulation BWR dynamics with stochastic and deterministic methods

    International Nuclear Information System (INIS)

    VanderHagen, T.H.; Van Dam, H.; Hoogenboom, J.E.; Kleiss, E.B.J.; Nissen, W.H.M.; Oosterkamp, W.J.

    1986-01-01

    Reactor kinetic, thermal hydraulic and total plant stability of a natural convection cooled BWR was studied using noise analysis and by evaluation of process responses to control rod steps and to steamflow control valve steps. An estimate of the fuel thermal time constant and an impression of the recirculation flow response to power variations was obtained. A sophisticated noise analysis method resulted in more insight into the fluctuations of the coolant velocity

  5. Instability and star evolution

    International Nuclear Information System (INIS)

    Mirzoyan, L.V.

    1981-01-01

    The observational data are discussed which testify that the phenomena of dynamical instability of stars and stellar systems are definite manifestations of their evolution. The study of these phenomena has shown that the instability is a regular phase of stellar evolution. It has resulted in the recognition of the most important regularities of the process of star formation concerning its nature. This became possible due to the discovery in 1947 of stellar associations in our Galaxy. The results of the study of the dynamical instability of stellar associations contradict the predictions of classical hypothesis of stellar condensation. These data supplied a basis for a new hypothesis on the formation of stars and nebulae by the decay of superdense protostars [ru

  6. Numerical study and modeling of hydrodynamic instabilities in the context of inertial confinement fusion in the presence of self-generated magnetic fields

    International Nuclear Information System (INIS)

    Levy, Y.

    2012-01-01

    In the context of inertial confinement fusion we investigate effects of magnetic fields on the development in the linear regime of two hydrodynamic instabilities: Richtmyer-Meshkov instability using ideal magnetohydrodynamics and ablative Rayleigh-Taylor instability in both acceleration and deceleration stages. Direct numerical simulations with a linear perturbation code enable us to confirm the stabilizing effect of the component of the magnetic field along the perturbations wave vector. The amplitude doesn't grow linearly in time but experiences oscillations instead. The compressibility taken into account in the code does not affect predictions given by an already existing impulsive and incompressible model. As far as Rayleigh-Taylor instability is concerned we study the effects of self-generated magnetic fields that arise from the development of the instability itself. In the acceleration stage we perform two dimensional simulations in planar geometry. We show that magnetic fields of about 1 T can be generated and that the instability growth transits more rapidly into nonlinear growth with the enhancement of the development of the third harmonic. We also propose an adaptation of an existing model that aims at studying thermal conductivity anisotropy effects, to take into account the effects of the self-generated magnetic fields on the Rayleigh-Taylor instability growth rate. Finally, in the deceleration stage, we perform two dimensional simulations in cylindrical geometry that take into account self-generation of magnetic fields due to the instability development. It reveals magnetic fields of about several thousands of Teslas that are not strong enough though to affect the instability behavior. (author) [fr

  7. European BWR R and D cluster for innovative passive safety systems

    International Nuclear Information System (INIS)

    Hicken, E.F.; Lensa, W. von

    1996-01-01

    The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de

  8. Fission product model for BWR analysis with improved accuracy in high burnup

    International Nuclear Information System (INIS)

    Ikehara, Tadashi; Yamamoto, Munenari; Ando, Yoshihira

    1998-01-01

    A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1%Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation. (author)

  9. IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1996-01-01

    Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

  10. Application of eddy current inspection to the Inconel weld of BWR internals

    International Nuclear Information System (INIS)

    Machida, Eiji; Yusa, Noritaka

    2004-01-01

    In order to definite the basic specifications of application of ECT (Eddy Current Test) to Inconel weld of BWR internals, the inspection and numerical analysis were carried out. The characteristics of the existing ECT probe were studied by making sample as same as CRD stud tube, measuring the relative permeability and electric conductivity of Inconel and alloy and evaluating ECT probe. On the basis of the results obtained, the basic specifications were determined and a new eddy current probe for inspection was designed and produced. The new ECT probe was able to detect small notch in Inconel weld, to classify the defects by eddy current inspection signal and sizing the length and depth. It is concluded that the new ECT probe is able to apply the Inconel weld of BWR internals. (S.Y.)

  11. An experimental study of two-phase flow instability on two parallel channel with low steam quality

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu shaorong; Bo Jinhai; Yao Meisheng; Han Bing; Zhang Youjie

    1988-01-01

    An experimental result of two-phase flow instability on two parallel channel natural circulation with low steam quality is presented. The comparison of instability in the single channel and that in parallel channel is given. The effect of unequal inlet resistance coefficient and unequal power on the parallel channel instability is described and the behaviour of instability with equal exit steam quality in the two channel is investigated

  12. Experimental study on ablative stabilization of Rayleigh-Taylor instability of laser-irradiated targets

    Science.gov (United States)

    Shigemori, Keisuke; Sakaiya, Tatsuhiko; Otani, Kazuto; Fujioka, Shinsuke; Nakai, Mitsuo; Azechi, Hiroshi; Shiraga, Hiroyuki; Tamari, Yohei; Okuno, Kazuki; Sunahara, Atsushi; Nagatomo, Hideo; Murakami, Masakatsu; Nishihara, Katsunobu; Izawa, Yasukazu

    2004-09-01

    Hydrodynamic instabilities are key issues of the physics of inertial confinement fusion (ICF) targets. Among the instabilities, Rayleigh-Taylor (RT) instability is the most important because it gives the largest growth factor in the ICF targets. Perturbations on the laser irradiated surface grow exponentially, but the growth rate is reduced by ablation flow. The growth rate γ is written as Takabe-Betti formula: γ = [kg/(1+kL)]1/2-βkm/pa, where k is wave number of the perturbation, g is acceleration, L is density scale-length, β is a coefficient, m is mass ablation rate per unit surface, and ρa is density at the ablation front. We experimentally measured all the parameters in the formula for polystyrene (CH) targets. Experiments were done on the HIPER laser facility at Institute of Laser Engineering, Osaka University. We found that the β value in the formula is ~ 1.7, which is in good agreements with the theoretical prediction, whereas the β for certain perturbation wavelengths are larger than the prediction. This disagreement between the experiment and the theory is mainly due to the deformation of the cutoff surface, which is created by non-uniform ablation flow from the ablation surface. We also found that high-Z doped plastic targets have multiablation structure, which can reduce the RT growth rate. When a low-Z target with high-Z dopant is irradiated by laser, radiation due to the high-Z dopant creates secondary ablation front deep inside the target. Since, the secondary ablation front is ablated by x-rays, the mass ablation rate is larger than the laser-irradiated ablation surface, that is, further reduction of the RT growth is expected. We measured the RT growth rate of Br-doped polystyrene targets. The experimental results indicate that of the CHBr targets show significantly small growth rate, which is very good news for the design of the ICF targets.

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

    International Nuclear Information System (INIS)

    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

  14. Study and mitigation of calibration factor instabilities in a water vapor Raman lidar

    Directory of Open Access Journals (Sweden)

    L. David

    2017-07-01

    Full Text Available We have investigated calibration variations in the Rameau water vapor Raman lidar. This lidar system was developed by the Institut National de l'Information Géographique et Forestière (IGN together with the Laboratoire Atmosphères, Milieux, Observations Spatiales (LATMOS. It aims at calibrating Global Navigation Satellite System (GNSS measurements for tropospheric wet delays and sounding the water vapor variability in the lower troposphere. The Rameau system demonstrated good capacity in retrieving water vapor mixing ratio (WVMR profiles accurately in several campaigns. However, systematic short-term and long-term variations in the lidar calibration factor pointed to persistent instabilities. A careful testing of each subsystem independently revealed that these instabilities are mainly induced by mode fluctuations in the optic fiber used to couple the telescope to the detection subsystem and by the spatial nonuniformity of the photomultiplier photocathodes. Laboratory tests that replicate and quantify these instability sources are presented. A redesign of the detection subsystem is presented, which, combined with careful alignment procedures, is shown to significantly reduce the instabilities. Outdoor measurements were performed over a period of 5 months to check the stability of the modified lidar system. The calibration changes in the detection subsystem were monitored with lidar profile measurements using a common nitrogen filter in both Raman channels. A short-term stability of 2–3 % and a long-term drift of 2–3 % per month are demonstrated. Compared to the earlier Development of Methodologies for Water Vapour Measurement (DEMEVAP campaign, this is a 3-fold improvement in the long-term stability of the detection subsystem. The overall water vapor calibration factors were determined and monitored with capacitive humidity sensor measurements and with GPS zenith wet delay (ZWD data. The changes in the water vapor calibration factors

  15. Tests of a numerical algorithm for the linear instability study of flows on a sphere

    Energy Technology Data Exchange (ETDEWEB)

    Perez Garcia, Ismael; Skiba, Yuri N [Univerisidad Nacional Autonoma de Mexico, Mexico, D.F. (Mexico)

    2001-04-01

    A numerical algorithm for the normal mode instability of a steady nondivergent flow on a rotating sphere is developed. The algorithm accuracy is tested with zonal solutions of the nonlinear barotropic vorticity equation (Legendre polynomials, zonal Rossby-Harwitz waves and monopole modons). [Spanish] Ha sido desarrollado un algoritmo numerico para estudiar la inestabilidad lineal de un flujo estacionario no divergente en una esfera en rotacion. La precision del algoritmo se prueba con soluciones zonales de la ecuacion no lineal de vorticidad barotropica (polinomios de Legendre, ondas zonales Rossby-Harwitz y modones monopolares).

  16. Systematic study of the short-term instability of PbWO4 scintillator parameters under irradiation

    International Nuclear Information System (INIS)

    Annenkov, A.N.; Ligun, A.B.; Auffray, E.; Lecoq, P.; Chipaux, R.; Geleoc, M.; Drobychev, G. Yu; Fedorov, A.A.; Korzhik, M.V.; Missevitch, O.V.; Pavlenko, V.B.; Golubev, N.A.; Lednev, A.A.; Peigneux, J.-P.; Singovski, A.V.

    1998-01-01

    The effect of irradiation on lead tungstate PbWO 4 (PWO) scintillator properties has been studied at different irradiation facilities. Lead tungstate crystals, grown with the oxide content in the melt tuned to the stoichiometry of pure sheelite or sheelite-like crystal types and doped with heterovalent, trivalent and pentavalent impurities, have been studied in order to optimize their resistance to irradiation. A combination of a selective cleaning of raw materials, a tuning of the melt from crystallization to crystallization and a destruction or compensation of the point-structure defects has to be used to minimize the short-term instability of PWO parameters under irradiation

  17. ECP measurements in the BWR-1 water loop relative to water composition changes

    Energy Technology Data Exchange (ETDEWEB)

    Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2010-07-01

    The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

  18. Assessment of radioactive residues arising from radiolabel instability in a multiple dose tissue distribution study in rats

    Energy Technology Data Exchange (ETDEWEB)

    Slatter, J.G. [Pharmacia Corp., Peapack, NJ (United States); Sams, J.P.; Easter, J.A. [Pharmacia Corp., Kalamazoo, MI (United States)] [and others

    2003-05-01

    Our study objectives were to quantitatively determine the effect of radiolabel instability on terminal phase radioactive tissue residues in a multiple dose tissue distribution study, to quantitatively compare tissue residue artifacts (non drug-related radioactivity) from two chemically-distinct radiolabel locations, and to conduct a definitive multiple dose tissue distribution study using the better of the two radiolabeled compounds. We compared the excretion and tissue distribution in rats of [{sup 14}C]linezolid, radiolabeled in two different locations, after 7 consecutive once daily [{sup 14}C] oral doses. The radiolabels were in the acetamide (two carbon) and oxazolidinone (isolated carbon) functional groups. Terminal phase tissue residue and excretion data were compared to data from rats dosed orally with [{sup 14}C]sodium acetate. Drug-related radioactivity was excreted rapidly over 24 h. After a single dose, the acetamide and oxazolidinone radiolabel sites both gave 3% of dose as exhaled {sup 14}CO{sub 2}. After 7 daily [{sup 14}C] oral doses, terminal phase radioactive tissue residues were higher from the acetamide radiolabel, relative to the oxazolidinone radiolabel, and were primarily not drug-related. In the definitive tissue distribution study, low concentrations of drug-related radioactivity in skin and thyroid were observed. We conclude that although small amounts of radiolabel instability do not significantly affect single dose tissue radioactivity C{sub max} and area under the curve (AUC), artifacts arising from radiolabel instability can prolong the apparent terminal phase half life and complicate study data interpretation. When possible, it is always preferable to use a completely stable radiolabel site. (author)

  19. Assessment of radioactive residues arising from radiolabel instability in a multiple dose tissue distribution study in rats

    International Nuclear Information System (INIS)

    Slatter, J.G.; Sams, J.P.; Easter, J.A.

    2003-01-01

    Our study objectives were to quantitatively determine the effect of radiolabel instability on terminal phase radioactive tissue residues in a multiple dose tissue distribution study, to quantitatively compare tissue residue artifacts (non drug-related radioactivity) from two chemically-distinct radiolabel locations, and to conduct a definitive multiple dose tissue distribution study using the better of the two radiolabeled compounds. We compared the excretion and tissue distribution in rats of [ 14 C]linezolid, radiolabeled in two different locations, after 7 consecutive once daily [ 14 C] oral doses. The radiolabels were in the acetamide (two carbon) and oxazolidinone (isolated carbon) functional groups. Terminal phase tissue residue and excretion data were compared to data from rats dosed orally with [ 14 C]sodium acetate. Drug-related radioactivity was excreted rapidly over 24 h. After a single dose, the acetamide and oxazolidinone radiolabel sites both gave 3% of dose as exhaled 14 CO 2 . After 7 daily [ 14 C] oral doses, terminal phase radioactive tissue residues were higher from the acetamide radiolabel, relative to the oxazolidinone radiolabel, and were primarily not drug-related. In the definitive tissue distribution study, low concentrations of drug-related radioactivity in skin and thyroid were observed. We conclude that although small amounts of radiolabel instability do not significantly affect single dose tissue radioactivity C max and area under the curve (AUC), artifacts arising from radiolabel instability can prolong the apparent terminal phase half life and complicate study data interpretation. When possible, it is always preferable to use a completely stable radiolabel site. (author)

  20. Instability characteristics of fluidelastic instability of tube rows in crossflow

    International Nuclear Information System (INIS)

    Chen, S.S.; Jendrzejczyk, J.A.

    1986-04-01

    An experimental study is reported to investigate the jump phenomenon in critical flow velocities for tube rows with different pitch-to-diameter ratios and the excited and intrinsic instabilities for a tube row with a pitch-to-diameter ratio of 1.75. The experimental data provide additional insights into the instability phenomena of tube arrays in crossflow. 9 refs., 10 figs

  1. Fingerprints of dynamical instabilities

    International Nuclear Information System (INIS)

    Chomaz, Ph.; Colonna, M.; Guarnera, A.

    1993-01-01

    It is explained why any reduced descriptions, such as mean field approximation, are stochastic in nature. It is shown that the introduction of this stochastic dynamics leads to a predictive theory in a statistical sens whatever the individual trajectories are characterized by the occurrence of bifurcations, instabilities or phase transitions. Concerning nuclear matter, the spinodal instability is discussed. In such a critical situation, the possibility to replace the stochastic part of the collision integral in the Boltzmann-Langevin model by the numerical noise associated with the finite number of test particles in ordinary BUU treatment is studied. It is shown that the fingerprints of these instabilities are kept during the evolution because of the relatively long recombination time compared with the typical time scales imposed by the Coulomb repulsion and the possible collective expansion. (author) 5 refs., 12 figs

  2. Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

    1986-01-01

    The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

  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. Causes of genome instability

    DEFF Research Database (Denmark)

    Langie, Sabine A S; Koppen, Gudrun; Desaulniers, Daniel

    2015-01-01

    function, chromosome segregation, telomere length). The purpose of this review is to describe the crucial aspects of genome instability, to outline the ways in which environmental chemicals can affect this cancer hallmark and to identify candidate chemicals for further study. The overall aim is to make......Genome instability is a prerequisite for the development of cancer. It occurs when genome maintenance systems fail to safeguard the genome's integrity, whether as a consequence of inherited defects or induced via exposure to environmental agents (chemicals, biological agents and radiation). Thus...

  5. Instabilities and nonequilibrium structures

    International Nuclear Information System (INIS)

    Tirapegui, E.; Villarroel, D.

    1987-01-01

    Physical systems can be studied both near to and far from equilibrium where instabilities appear. The behaviour in these two regions is reviewed in this book, from both the theoretical and application points of view. The influence of noise in these situations is an essential feature which cannot be ignored. It is therefore discussed using phenomenological and theoretical approaches for the numerous problems which still remain in the field. This volume should appeal to mathematicians and physicists interested in the areas of instability, bifurcation theory, dynamical systems, pattern formation, nonequilibrium structures and statistical mechanics. (Auth.)

  6. Simulation studies of plasma waves in the electron foreshock: The transition from reactive to kinetic instability

    International Nuclear Information System (INIS)

    Dum, C.T.

    1990-01-01

    The electron beam-plasma instability is analyzed in particle simulation experiments, starting with a beam of small velocity spread. The dispersion relation is solved for snapshots of the actual evolving electron distribution function, rather than for the usual models consisting of Maxwellians. As the beam broadens, the analysis shows a transition from reactive beam modes, with frequencies extending much below the plasma frequency ω e , to kinetic instability of Langmuir waves, ω∼ω e , which is in agreement with the frequencies and growth rates observed in the simulation. Beam evolution is also in agreement with quasi-linear theory, except at the end of the reactive phase when trapping of beam electrons is seen. Although the spectrum temporarily narrows at this stage, there are, in contrast to previous simulations, still many modes present. the system then can proceed to a kinetic phase in which quasi-linear theory is again applicable. This stage is identical with the evolution starting from a gentle broad beam, except that wave levels are several times higher. With higher wave levels, mode coupling effects are also more prominent, but are still unable to prevent plateau formation. In contrast to the Langmuir wave regime, the reactive broadband wave regime lasts only for a relatively short period. In the electron foreshock it could only persist if a narrow beam or a sharp cutoff feature were maintained by continued beam injection and the time-of-flight mechanism

  7. Study of microwave instabilities by means of a square-well potential

    International Nuclear Information System (INIS)

    Kim, K.J.

    1979-01-01

    Microwave instabilities are analyzed in a simple model, in which the usual synchrotron oscillation of a particle is replaced by particle motion in a square-well potential. In the usual synchrotron oscillation, a particle moves along an elliptic trajectory. The most natural coordinates for such a motion are the action and the angle variables. On the other hand, the distribution of the particles along the ring is most conveniently described by azimuthal variables. The difficulty disappears if the synchrotron motion is approximated by the motion in a square-well potential. The square-well potential may seem extremely unphysical. However, it should be remarked that the form of the potential with addition of a Landau cavity looks more or less like a square-well. At any rate, the main motivation of introducing the square-well here is to simplify the mathematics of and thereby gain some insight into microwave instabilities. The model is exactly soluble. The results are in general agreement with the conclusions obtained from qualitative arguments based on coasting beam theory. However, some of the detailed features of the solution, for example the behavior of ω 2 as a function of impedance, are surprising

  8. A Numerical Study of Nonlinear Nonhydrostatic Conditional Symmetric Instability in a Convectively Unstable Atmosphere.

    Science.gov (United States)

    Seman, Charles J.

    1994-06-01

    Nonlinear nonhydrostatic conditional symmetric instability (CSI) is studied as an initial value problem using a two-dimensional (y, z)nonlinear, nonhydrostatic numerical mesoscale/cloud model. The initial atmosphere for the rotating, baroclinic (BCF) simulation contains large convective available potential energy (CAPE). Analytical theory, various model output diagnostics, and a companion nonrotating barotropic (BTNF) simulation are used to interpret the results from the BCF simulation. A single warm moist thermal initiates convection for the two 8-h simulations.The BCF simulation exhibited a very intricate life cycle. Following the initial convection, a series of discrete convective cells developed within a growing mesoscale circulation. Between hours 4 and 8, the circulation grew upscale into a structure resembling that of a squall-line mesoscale convective system (MCS). The mesoscale updrafts were nearly vertical and the circulation was strongest on the baroclinically cool side of the initial convection, as predicted by a two-dimensional Lagrangian parcel model of CSI with CAPE. The cool-side mesoscale circulation grew nearly exponentially over the last 5 h as it slowly propagated toward the warm air. Significant vertical transport of zonal momentum occurred in the (multicellular) convection that developed, resulting in local subgeostrophic zonal wind anomalies aloft. Over time, geostrophic adjustment acted to balance these anomalies. The system became warm core, with mesohigh pressure aloft and mesolow pressure at the surface. A positive zonal wind anomaly also formed downstream from the mesohigh.Analysis of the BCF simulation showed that convective momentum transport played a key role in the evolution of the simulated MCS, in that it fostered the development of the nonlinear CSI on mesoscale time scales. The vertical momentum transport in the initial deep convection generated a subgeostrophic zonal momentum anomaly aloft; the resulting imbalance in pressure

  9. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-01-01

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  10. Saturation of equatorial inertial instability

    NARCIS (Netherlands)

    Kloosterziel, R.C.; Orlandi, P.; Carnevale, G.F.

    2015-01-01

    Inertial instability in parallel shear flows and circular vortices in a uniformly rotating system ( $f$f-plane) redistributes absolute linear momentum or absolute angular momentum in such a way as to neutralize the instability. In previous studies we showed that, in the absence of other

  11. Product Evaluation Task Force Phase Two report for BWR/PWR dissolver wastes

    International Nuclear Information System (INIS)

    Francis, A.J.

    1990-01-01

    It has been proposed that all Intermediate Level Wastes arising at Sellafield should be encapsulated prior to ultimate disposal. The Product Evaluation Task Force (PETF) was set up to investigate possible encapsulants and to produce an adequate data base to justify the preferred matrices. This report details the work carried out, under Phase 2 of the Product Evaluation Task Force programme, on BWR/PWR Dissolver Wastes. Three possible types of encapsulants for BWR/PWR Dissolver Wastes:- Inorganic cements, Polymer cements and Polymers are evaluated using the Kepner Tregoe decision analysis technique. This technique provides a methodology for scoring and ranking alternative options and evaluating any risks associated with an option. The analysis shows that for all four stages of waste management operations ie Storage, Transport, handling and emplacement, Disposal and Process, cement matrices are considerably superior to other potential matrices. A matrix, consisting of three parts Blast Furnace Slag (BFS) to one part Ordinary Portland Cement (OPC), is recommended for Phase 3 studies on BWR/PWR Dissolver Wastes. (author)

  12. Role of BWR secondary containments in severe accident mitigation: issues and insights from recent analyses

    International Nuclear Information System (INIS)

    Greene, S.R.

    1988-01-01

    All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by a secondary containment consisting of a reactor building and refueling bay (MK I and MK II designs), a shield building, auxiliary building and fuel building (MK III), or an auxiliary building and enclosure building (Grand Gulf style MK III). Although secondary containment designs are highly plant specific, their purpose is to minimize the ground level release of radioactive material for a spectrum of traditional design basis accidents. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident. This paper presents a brief overview of domestic BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent ORNL secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented

  13. Stability monitoring for BWR based on singular value decomposition method using artificial neural network

    International Nuclear Information System (INIS)

    Tsuji, Masashi; Shimazu, Yoichiro; Michishita, Hiroshi

    2005-01-01

    A new method for evaluating the decay ratios in a boiling water reactor (BWR) using the singular value decomposition (SVD) method had been proposed. In this method, a signal component closely related to the BWR stability can be extracted from independent components of the neutron noise signal decomposed by the SVD method. However, real-time stability monitoring by the SVD method requires an efficient procedure for screening such components. For efficient screening, an artificial neural network (ANN) with three layers was adopted. The trained ANN was actually applied to decomposed components of local power range monitor (LPRM) signals that were measured in stability experiments conducted in the Ringhals-1 BWR. In each LPRM signal, multiple candidates were screened from the decomposed components. However, decay ratios could be estimated by introducing appropriate criterions for selecting the most suitable component among the candidates. The estimated decay ratios are almost identical to those evaluated by visual screening in a previous study. The selected components commonly have the largest singular value, the largest decay ratio and the least squared fitting error among the candidates. By virtue of excellent screening performance of the trained ANN, the real-time stability monitoring by the SVD method can be applied in practice. (author)

  14. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  15. Analysis CFD for the hydrogen transport in the primary containment of a BWR

    International Nuclear Information System (INIS)

    Jimenez P, D. A.; Del Valle G, E.; Gomez T, A. M.

    2014-10-01

    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)

  16. Development status of compact containment BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

    2005-01-01

    In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response

  17. Comparison of thoracolumbar motion produced by manual and Jackson-table-turning methods. Study of a cadaveric instability model.

    Science.gov (United States)

    DiPaola, Christian P; DiPaola, Matthew J; Conrad, Bryan P; Horodyski, MaryBeth; Del Rossi, Gianluca; Sawers, Andrew; Rechtine, Glenn R

    2008-08-01

    Patients who have sustained a spinal cord injury remain at risk for further neurologic deterioration until the spine is adequately stabilized. To our knowledge, no study has previously addressed the effects of different bed-to-operating room table transfer techniques on thoracolumbar spinal motion in an instability model. We hypothesized that the conventional logroll technique used to transfer patients from a supine position to a prone position on the operating room table has the potential to confer significantly more motion to the unstable thoracolumbar spine than the Jackson technique. Three-column instability was surgically created at the L1 level in seven cadavers. Two protocols were tested. The manual technique entailed performing a standard logroll of a supine cadaver to a prone position on an operating room Jackson table. The Jackson technique involved sliding the supine cadaver to the Jackson table, securing it to the table, and then rotating it into a prone position. An electromagnetic tracking device measured motion--i.e., angular motion (flexion-extension, lateral bending, and axial rotation) and linear translation (axial, medial-lateral, and anterior-posterior) between T12 and L2. The logroll technique created significantly more motion than the Jackson technique as measured with all six parameters. Manual logroll transfers produced an average of 13.8 degrees to 18.1 degrees of maximum angular displacement and 16.6 to 28.3 mm of maximum linear translation. The Jackson technique resulted in an average of 3.1 degrees to 5.8 degrees of maximum angular displacement (p patient safety. Performing the Jackson turn requires approximately half as many people as required for a manual logroll. This study suggests that the Jackson technique should be considered for supine-to-prone transfer of patients with known or suspected instability of the thoracolumbar spine.

  18. Elbow joint instability

    DEFF Research Database (Denmark)

    Olsen, Bo Sanderhoff; Henriksen, M G; Søjbjerg, Jens Ole

    1994-01-01

    The effect of simultaneous ulnar and radial collateral ligament division on the kinematics of the elbow joint is studied in a cadaveric model. Severance of the anterior part of the ulnar collateral ligament and the annular ligament led to significant elbow joint instability in valgus and varus...

  19. Study of tertiary creep instability in several elevated-temperature structural materials

    International Nuclear Information System (INIS)

    Booker, M.K.; Sikka, V.K.

    1978-01-01

    Data for a number of common elevated temperature structural materials have been analyzed to yield mathematical predictions for the time and strain to tertiary creep at various rupture lives and temperatures. Materials examined include types 304 and 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, alloy 718, Hastelloy alloy X, and ERNiCr--3 weld metal. Data were typically examined over a range of creep temperatures for rupture lives ranging from less than 100 to greater than 10,000 hours. Within a given material, trends in these quantities can be consistently described, but it is difficult to directly relate the onset of tertiary creep to failure-inducing instabilities. A series of discontinued tests for alloy 718 at 649 and 620 0 C showed that the material fails by intergranular cracking but that no significant intergranular cracking occurs until well after the onset of tertiary creep

  20. Use of one-dimensional Cosserat theory to study instability in a viscous liquid jet

    International Nuclear Information System (INIS)

    Bogy, D.B.

    1978-01-01

    The problem of the instability of an incompressible viscous liquid jet is considered within the context of one-dimensional Cosserat equations. Linear stability analyses are performed for both the infinite and semi-infinite jets. The results obtained for the inviscid case are compared with the corresponding results derived from ideal fluid equations. They are also compared with recent results by other authors obtained from a different set of one-dimensional jet equations. Solutions are also obtained, within the framework of the linearized theory, to the jet break-up problems formulated as an initial-value problem for the infinite jet and as a boundary-value problem for the semi-infinite jet