Reactor noise in critical and accelerator driven sub-critical systems

International Nuclear Information System (INIS)

Degweker, S.B.; Rana, Y.S.

2007-01-01

Noise methods have long been used for reactor kinetics parameters measurement and as diagnostic tools for monitoring the health of a nuclear power plant. It is conceivable that noise techniques would find similar applications in ADS. Measurement/monitoring the degree of sub-criticality of an ADS is one such application for which noise based methods are being considered, among others such as the pulsed source method. For this reason, theoretical studies on ADS noise have appeared since the late nineties. The principal difference between critical reactor noise and ADS noise is due to the statistical properties of the source. Unlike the source due to radioactive decay present in ordinary reactors, the machine produced ADS source cannot be assumed to be a Poisson process. In addition the source is pulsed. All this requires a new theoretical approach to the subject. In a number of papers (beginning in 2000) such a theoretical approach has been developed in BARC. Over the years, our approach has received general acceptance. The paper gives a description of the subject of reactor noise and its applications in critical reactors. The theory of noise in ADS is then outlined, highlighting the differences in approach and results from that of critical reactors. (author)

Reactor noise monitoring device

International Nuclear Information System (INIS)

Yamanaka, Hiroto.

1990-01-01

The present invention concerns a reactor noise monitoring device by detecting abnormal sounds in background noises. Vibration sounds detected by accelerometers are applied to a loose parts detector. The detector generates high alarm if there are sudden impact sounds in the background noises and applies output signals to an accumulation device. If there is slight impact sounds in the vicinity of any of the accelerometers, the accumulation device accumulates the abnormal sounds assumed to be generated from an identical site while synchronizing the waveforms for all of the channels. Then, the device outputs signals in which the background noises are cancelled, as detection signals. Therefore, S/N ratio can be improved and the abnormal sounds contained in the background noises can be detected, to thereby improve the accuracy for estimating the position where the abnormal sounds are generated. (I.S.)

Frequency spectrum of Calder Hall reactor noise

International Nuclear Information System (INIS)

Cummins, J.D.

1960-01-01

The frequency spectrum of the noise power of Calder Hall reactor No. 1 has been obtained by analysing a tape recording of the backed off power. The root mean square noise power due to all frequencies above 0.001 cycles per second was found to be 0.13%. The noise power for this reactor, is due mainly to modulations of the power level by reactivity variations caused in turn by gas temperature changes. These gas temperature changes are caused by a Cyclic variation in the feedwater regulator to the heat exchanger. The apparatus and method used to determine the noise power are described in this memorandum. It is shown that for frequencies in the range 0.001 to 0.030 cycles per second the noise spectrum falls at 60 decibels per decade of frequency. (author)

Reactor noise analysis applications in NPP I and C systems

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

2006-07-01

Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

A new neutron noise technique for fast reactors

International Nuclear Information System (INIS)

Zhuo Fengguan; Jin Manyi; Yao Shigui; Su Zhuting

1987-12-01

This paper gives a new neutron noise technique for fast reactors, which is known as thermalization measurement technique of the neutron noise. The theoretical formulas of the technique were developed, and a digital delayed coincidence time analyzer consisted of TTL integrated circuits was constructed for the study of this technique. The technique has been tested and applied practically at Df-VI fast zero power reactor. It was shown that the provided technique in this work has a number of significant advantages in comparison with the conventional neutron noise method

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

A multi-group neutron noise simulator for fast reactors

International Nuclear Information System (INIS)

Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

2013-01-01

Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

Recent progress in safety-related applications of reactor noise analysis

International Nuclear Information System (INIS)

Hirota, Jitsuya; Shinohara, Yoshikuni; Saito, Keiichi

1982-01-01

Recent progress in safety-related applications of reactor noise analysis is reviewed, mainly referring to various papers presented at the Third Specialists' Meeting on Reactor Noise (SMORN-III) held in Tokyo in 1981. Advances in application of autoregressive model, coherence analysis and pattern recognition technique are significant since SMORN-II in 1977. Development of reactor diagnosis systems based on noise analysis is in progress. Practical experiences in the safety-related applications to power plants are being accumulated. Advances in quantitative monitoring of vibration of internal structures in PWR and diagnosis of core stability and control system characteristics in BWR are notable. Acoustic methods are also improved to detect sodium boiling in LMFBR. The Reactor Noise Analysis Benchmark Test performed by Japan in connection with SMORN-III is successful so that it is possible to proceed to the second stage of the benchmark test. (author)

Advances in automated noise data acquisition and noise source modeling for power reactors

International Nuclear Information System (INIS)

Clapp, N.E. Jr.; Kryter, R.C.; Sweeney, F.J.; Renier, J.A.

1981-01-01

A newly expanded program, directed toward achieving a better appreciation of both the strengths and limitations of on-line, noise-based, long-term surveillance programs for nuclear reactors, is described. Initial results in the complementary experimental (acquisition and automated screening of noise signatures) and theoretical (stochastic modeling of likely noise sources) areas of investigation are given

Study on statistical analysis of nonlinear and nonstationary reactor noises

International Nuclear Information System (INIS)

Hayashi, Koji

1993-03-01

For the purpose of identification of nonlinear mechanism and diagnosis of nuclear reactor systems, analysis methods for nonlinear reactor noise have been studied. By adding newly developed approximate response function to GMDH, a conventional nonlinear identification method, a useful method for nonlinear spectral analysis and identification of nonlinear mechanism has been established. Measurement experiment and analysis were performed on the reactor power oscillation observed in the NSRR installed at the JAERI and the cause of the instability was clarified. Furthermore, the analysis and data recording methods for nonstationary noise have been studied. By improving the time resolution of instantaneous autoregressive spectrum, a method for monitoring and diagnosis of operational status of nuclear reactor has been established. A preprocessing system for recording of nonstationary reactor noise was developed and its usability was demonstrated through a measurement experiment. (author) 139 refs

Analysis of log rate noise in Ontario's CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Hinds, H.W. [Dynamic Simulation and Analysis Corp., Deep River, Ontario (Canada); Banica, C.; Arguner, D. [Ontario Power Generation, Ajax, Ontario (Canada); Scharfenberg, R. [Bruce Power, Tiverton, Ontario (Canada)

2007-07-01

In the fall of 2003, the operators noticed that in the recently-refurbished Bruce A Shutdown System no. 1 (SDS1) the noise level in Log Rate signals were much larger than before. At the request of the Canadian Nuclear Safety Commission (CNSC), all Canadian CANDU reactors took action to characterize their Log Rate noise. Staff of the Inspection and Maintenance Services division of Ontario Power Generation (OPG) has collected high-speed high-accuracy noise data from nearly all 16 Ontario reactors, either as part of routine measurements before planned outages or as a dedicated noise recording. This paper gives the results of examining a suitable subset of this data, with respect to the characteristics and possible causes of Log Rate noise. The reactor and instrumentation design is different at each station: the locations of the moderator injection nozzles, the location of the ion chambers for each system, and the design of the Log Rate amplifiers. It was found that the Log noise (source of Log Rate noise) was much larger for those ion chambers in the path of the moderator injection nozzles, compared to those which were not in the path. This 'extra' Log noise would then be either attenuated or amplified depending on the transfer function (time constants) of the Log Rate amplifier. It was also observed that most of the Log and Log Rate noise is independent of any other signal measured. Although all CANDU reactors in Ontario have Log and Log Rate noise, the Bruce A SDS1 system has the largest amount of Log Rate noise, because (a) its SDS1 (and RRS) ion chambers are at the top of the reactor in the path of the moderator injection nozzles, and (b) its SDS1 Log Rate amplifiers have the smallest time constants. (author)

Neutron noise analysis for malfunction diagnosis at sodium cooled reactors

International Nuclear Information System (INIS)

Hoppe, P.

1978-09-01

For the investigation of the potential use of neutron noise analysis at sodium cooled power reactors, measurements have been performed at the KNK I reactor over a period of 18 month under different operational conditions. The signal fluctuations of the following tranducers have been recorded: In-core and Ex-core neutron detectors, temperature-, flow-, pressure-, vibration- and acoustic sensors. These extensive measurements have been analyzed in the frequency range from 0,001 Hz to 1000 Hz with all currently known methods for the identification of noise sources. The following results have been found: - Neutron noise for f 20 Hz the white detection noise prevails. In the region from 1 Hz to 20 Hz the vibrations of core components contribute to neutron noise. - Neutron noise is influenced by the state of the plant. - The contributions to neutron noise due to the fluctuations of coolant flow and inlet temperature are small compared to those produced by the movements of the control rod initiated by the reactor control system. The quantitatively unidentifiable amount of reactivity fluctuations (0,6 time-dependent thermal bowing of the core. With respect to these results and by calculation of the neutron noise patterns to be expected for the SNR 300, the following possible applications for neutron noise analysis have been found: By means of neutron noise analysis only reactivity fluctuations can be identified and supervised which are produced by time dependent changes of the core geometry. Furthermore neutron noise analysis is well suited for a sensitive detection of control rod vibrations and of local sodium boiling. Finally it can be used for the surveillance of the proper functioning of the reactor control system and of the control rod drive mechanism. (orig./HP) 891 HP [de

Application of autoregressive moving average model in reactor noise analysis

International Nuclear Information System (INIS)

Tran Dinh Tri

1993-01-01

The application of an autoregressive (AR) model to estimating noise measurements has achieved many successes in reactor noise analysis in the last ten years. The physical processes that take place in the nuclear reactor, however, are described by an autoregressive moving average (ARMA) model rather than by an AR model. Consequently more correct results could be obtained by applying the ARMA model instead of the AR model to reactor noise analysis. In this paper the system of the generalised Yule-Walker equations is derived from the equation of an ARMA model, then a method for its solution is given. Numerical results show the applications of the method proposed. (author)

Noise analysis of the Dodewaard boiling water reactor: characteristics and time history

International Nuclear Information System (INIS)

Veer, J.H.C. v.d.; Kema, N.V.

1982-01-01

Reactor noise measurements have been performed in the Dodewaard BWR since the eighth fuel cycle (1978). Analysis of the noise characteristics of the ex-core neutron detectors are reported. As a result characteristics of the global component of the boiling noise and the influence of oscillatory effects in reactor pressure control and steam flow rate are described. The influence of power feedback effects on the detection of global noise at low frequencies is given using point kinetic reactor theory. Results are reported on the behaviour of the neutron noise characteristics during one fuel cycle and on the behaviour from fuel cycle 8 to 11. (author)

Investigations of certain problems in reactor noise diagnostics

International Nuclear Information System (INIS)

Karlsson, Joakim.

1997-06-01

A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The papers presented cover three relatively different areas, namely neutron noise induced by small vibrations of a strong absorber, the decomposition of noise in BWRs with application to stability monitoring, and fluctuation analysis of soft x-ray signals from fusion plasmas. 18 refs, 6 figs

Investigations of certain problems in reactor noise diagnostics

Energy Technology Data Exchange (ETDEWEB)

Karlsson, Joakim

1997-06-01

A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The papers presented cover three relatively different areas, namely neutron noise induced by small vibrations of a strong absorber, the decomposition of noise in BWRs with application to stability monitoring, and fluctuation analysis of soft x-ray signals from fusion plasmas. 18 refs, 6 figs.

Results of theoretical studies on power noise of PWR type reactors

International Nuclear Information System (INIS)

Meyer, K.

1979-12-01

For evaluating noise measurements of neutron flux in power reactors position-dependent effects are taken into account. Based on an effective one-group diffusion model results are obtained for reactor noise due to vibrating control elements and disturbances in the coolant flow

Characterization of noise sources in nuclear power reactors

International Nuclear Information System (INIS)

Andhill, Gustav

2004-03-01

Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D

Characterization of noise sources in nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Andhill, Gustav

2004-03-01

Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D.

Reactor noise diagnostics based on multivariate autoregressive modeling: Application to LOFT [Loss-of-Fluid-Test] reactor process noise

International Nuclear Information System (INIS)

Gloeckler, O.; Upadhyaya, B.R.

1987-01-01

Multivariate noise analysis of power reactor operating signals is useful for plant diagnostics, for isolating process and sensor anomalies, and for automated plant monitoring. In order to develop a reliable procedure, the previously established techniques for empirical modeling of fluctuation signals in power reactors have been improved. Application of the complete algorithm to operational data from the Loss-of-Fluid-Test (LOFT) Reactor showed that earlier conjectures (based on physical modeling) regarding the perturbation sources in a Pressurized Water Reactor (PWR) affecting coolant temperature and neutron power fluctuations can be systematically explained. This advanced methodology has important implication regarding plant diagnostics, and system or sensor anomaly isolation. 6 refs., 24 figs

Non-linearity consideration when analyzing reactor noise statistical characteristics. [BWR

Energy Technology Data Exchange (ETDEWEB)

Kebadze, B V; Adamovski, L A

1975-06-01

Statistical characteristics of boiling water reactor noise in the vicinity of stability threshold are studied. The reactor is considered as a non-linear system affected by random perturbations. To solve a non-linear problem the principle of statistical linearization is used. It is shown that the halfwidth of resonance peak in neutron power noise spectrum density as well as the reciprocal of noise dispersion, which are used in predicting a stable operation theshold, are different from zero both within and beyond the stability boundary the determination of which was based on linear criteria.

Identification of reactor failure states using noise methods, and spatial power distribution

International Nuclear Information System (INIS)

Vavrin, J.; Blazek, J.

1981-01-01

A survey is given of the results achieved. Methodical means and programs were developed for the control computer which may be used in noise diagnostics and in the control of reactor power distribution. Statistical methods of processing the noise components of the signals of measured variables were used for identifying failures of reactors. The method of the synthesis of the neutron flux was used for modelling and evaluating the reactor power distribution. For monitoring and controlling the power distribution a mathematical model of the reactor was constructed suitable for control computers. The uses of noise analysis methods are recommended and directions of further development shown. (J.P.)

Noise analysis role in reactor safety, Spectral analysis (PSD)

International Nuclear Information System (INIS)

Jovanovic, S.; Velickovic, Lj.

1967-11-01

Spectral power density of a zero power reactor is frequency dependent and related to transfer function of the reactor and to spectral density of the input disturbance. Measurement of spectral power density of a critical system is used to obtain the ratio (β/l), β is the effective yield of delayed neutrons, and l is the effective mean neutron lifetime. When reactor is subcritical, if the effective yie ald of delayed neutrons, the effective mean neutron lifetime are known, the shutdown margin can be determined by relation α = (1 - k (1- β0)/l, k is the effective multiplication factor. Output neutron spectrum at the RB reactor in Vinca was measured for a few reactor core configurations and for a few levels of heavy water at subcritical state. Measured values were satisfactory when the reactor was critical, but the reactor noise of subcritical system was covered by the white noise of the detector and electronic equipment. The Ra-Be source was under the reactor vessel when measurements of subcritical system were done. More efficient detector or external random stimulus for increasing the intensity of neutron fluctuations would be needed to obtain results for subcritical system

Physical model study of neutron noise induced by vibration of reactor internals

International Nuclear Information System (INIS)

Liu Jinhui; Gu Fangyu

1999-01-01

The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

Development and applications of reactor noise analysis at Ontario Hydro's CANDU reactors

International Nuclear Information System (INIS)

Gloeckler, O.; Tulett, M.V.

1995-01-01

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro's CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Fractal characterization for noise signal validation in power reactors

International Nuclear Information System (INIS)

Aguilar Martinez, Omar

2003-01-01

Up to now, a great variety of methods is used for the dynamical characterization of different components of Nuclear Power Plants (NPPs). With this aim, time and spectral analysis are usually considered, and different tools of non-stationary and non-gaussian analysis are also presented. When applying non-lineal dynamics theory for noise signal validation purposes in power reactors, the extraction of fractal echoes plays a main role. Fractal characterization for noise signal validation purposes can be integrated to the task of processing and acquisition of time signals in noise (fluctuation parameters) analysis systems. The possibility of discrimination between deterministic chaotic signals and pure noise signals has been incorporated, as a complement; to noise signals analysis in normal and anomalous operational conditions in NPPs using a fractal approach. In this work the detailed analysis of a neutronic sensor response is considered and the fractal characterization of its dynamics state (i.e. sensor line) for noise signal classification, it is presented. The experiment from where the time series (signals) were obtained, was carried out at the Research Reactor of the Technical University of Budapest, Hungary, during a model experiment for ageing process study of in-core neutron detectors (author)

New applications of neutron noise theory in power reactor physics

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, Vasiliy

2000-04-01

The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

New applications of neutron noise theory in power reactor physics

International Nuclear Information System (INIS)

Arzhanov, Vasiliy

2000-04-01

The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

Properties of autoregressive model in reactor noise analysis, 1

International Nuclear Information System (INIS)

Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

1987-01-01

Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

Power noise spectrum classification in the problem of the IBR-2 reactor

International Nuclear Information System (INIS)

Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

1988-01-01

The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

Localization of Vibrating Noise Sources in Nuclear Reactor Cores

International Nuclear Information System (INIS)

Hultqvist, Pontus

2004-09-01

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

Development and applications of reactor noise analysis at Ontario Hydro`s CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O [Ontario Hydro, Toronto, ON (Canada); Tulett, M V [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro`s CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Measurements of kinetic parameters by noise techniques on the MINERVE reactor

International Nuclear Information System (INIS)

Carre, J.C.; Da Costa Oliveira, J.

1975-01-01

Noise measurements were determined on ERMINE a fast thermal coupled reactor built in MINERVE. A reactor without feedback, and a reactor with an automatic control rod were both considered. The first case concerned the measurements of auto and cross power spectral density obtained with one or two neutron detectors, and the determination of: neutron lifetime; efficiency for one ion chamber; power level of the reactor; maximal speed and acceleration of the control rod for the design of an automatic reactor control actuator. The second case was concerned with measurements of the auto power spectral density in reactivity for the control rod, and the estimation of: the transfer function of the automatic pilot; the neutron lifetime; and the standard error affecting the results obtained by the oscillation method. The results proved that the pile noise theory with a point kinetic model is sufficient for application on zero power reactors. (U.K.)

Application of neutron noise analysis to a swimming pool research reactor

International Nuclear Information System (INIS)

Behringer, K.; Lescano, V.H.; Meier, F.; Phildius, J.; Winkler, H.

1982-01-01

This work is part of a programme of establishing practical applications of neutron noise techniques to a swimming pool research reactor and deals with two different items: (1) The identification of local boiling caused e.g. by a partial blockage of the coolant flow in a fuel element. Local boiling can easily lead to a burn-out situation. The onset of boiling can be detected by neutron noise analysis and a boiling detection system is presently under development. (2) The measurement of the time evolution of the reactivity induced by xenon after reactor shut-down by an on-line reactivity meter based on neutron noise analysis. From the data, the prompt neutron decay constant at delayed critical, the equilibrium xenon reactivity worth, and an estimate of the average steady-state power flux in the core before reactor shut-down were obtained. (author)

ARMA modeling of stochastic processes in nuclear reactor with significant detection noise

International Nuclear Information System (INIS)

Zavaljevski, N.

1992-01-01

The theoretical basis of ARMA modelling of stochastic processes in nuclear reactor was presented in a previous paper, neglecting observational noise. The identification of real reactor data indicated that in some experiments the detection noise is significant. Thus a more rigorous theoretical modelling of stochastic processes in nuclear reactor is performed. Starting from the fundamental stochastic differential equations of the Langevin type for the interaction of the detector with neutron field, a new theoretical ARMA model is developed. preliminary identification results confirm the theoretical expectations. (author)

Hydraulic noise in reactor circuits and loops, and its effect on nuclear fuel vibration

International Nuclear Information System (INIS)

Card, D.C.

This paper reports the results of an investigation at WNRE to monitor noise levels in reactor circuits and loops, so as to characterize the systems and establish the importance of this noise on fuel and pressure tube vibration. Some of the techniques necessary for in-reactor installations of pressure transducers have been developed and measurements have been obtained in the vertical fuel channels of a very noisy out-reactor loop as well as in the WR-1 reactor circuits. A very quiet out-reactor loop has been constructed to study the vibration behaviour of 37-element fuel bundles in the horizontal CANDU pressurized-heavy water reactor systems. In this facility various types and levels of hydraulic noise are being generated to study their effect on the fuel bundles and flow tube at flow velocities up to approximately 13 m/s. (author)

Data acquisition and processing system for reactor noise analysis

International Nuclear Information System (INIS)

Costa Oliveira, J.; Morais Da Veiga, C.; Forjaz Trigueiros, D.; Pombo Duarte, J.

1975-01-01

A data acquisition and processing system for reactor noise analysis by time correlation methods is described, consisting in one to four data feeding channels (transducer, associated electronics and V/f converter), a sampling unit, a landline transmission system and a PDP 15 computer. This system is being applied to study the kinetic parameters of the 'Reactor Portugues de Investigacao', a swimming-pool 1MW reactor. The main features that make such a data acquisition and processing system a useful tool to perform noise analysis are: the improved characteristics of analog-to-digital converters employed to quantize the signals; the use of an on-line computer which allows a great accumulation and a rapid treatment of data together with an easy check of the correctness of the experiments; and the adoption of the time cross-correlation technique using two-detectors which by-pass the limitation of low efficiency detectors. (author)

On the determination of boiling water reactor characteristics by noise analysis

International Nuclear Information System (INIS)

Kleiss, J.

1983-01-01

In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

Methods and tools to detect thermal noise in fast reactors

International Nuclear Information System (INIS)

Motta, M.; Giovannini, R.

1985-07-01

The Specialists' Meeting on ''Methods and Tools to Detect Thermal Noise in Fast Reactors'' was held in Bologna on 8-10 October 1984. The meeting was hosted by the ENEA and was sponsored by the IAEA on the recommendation of the International Working Group on Fast Reactors. 17 participants attended the meeting from France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, Joint Research Centre of CEC and from IAEA. The meeting was presided over by Prof. Mario Motta of Italy. The purpose of the meeting was to review and discuss methods and tools for temperature noise detection and related analysis as a potential means for detecting local blockages in fuel and blanket subassemblies and other faults in LMFBR. The meeting was divided into four technical sessions as follows: 1. National review presentations on application purposes and research activities for thermal noise detection. (5 papers); 2. Detection instruments and electronic equipment for temperature measurements in fast reactors. (5 papers); 3. Physical models. (2 papers); 4. Signal processing techniques. (3 papers). A separate abstract was prepared for each of these papers

Neutron noise measurement technique in a coupled reactor

International Nuclear Information System (INIS)

Genoud, J.P.

1976-01-01

Describes work carried out on the swimming pool reactor at the Physikalisch-Technische Bundesanstalt at Braunschweig. The reactor has two multiplying zones, is light water moderated, with 90% enriched 235 U fuel. There is a D 2 0 reservoir between the two parts of the reactor. Signal/noise ratio obtained by means of ionisation chamber type neutron detectors of 10 -13 amp/u.f. sensitivity is of the order of 40 dB and band frequency 1.5 kHz. Spectral density of the interzone interaction energy was obtained by use of Fourier transforms, previously corrected by a Hanning window. (S.W.)

Method of inspecting the function of reactor noise monitoring device

International Nuclear Information System (INIS)

Yamanaka, Hirohito.

1985-01-01

Purpose: To enable to inspect the function of a reactor noise monitoring device used for monitoring the operation abnormality in coolant circuits during reactor operation. Constitution: A cylinder incorporating a steel ball moved laterally by a pneumatic pressure is disposed to the main body of a reactor coolant circuit. A three-way solenoid valve disposed to a central control room outside to a radiation controlled area is connected with the cylinder by way of pneumatic pipeways. The three-way solenoid valve is operated for a certain period of time by a timer in the central control room to thereby impinge the steel ball in the cylinder against the main body of the coolant circuit and it is inspected as to whether the reactor noise monitoring system can detect the impinging energy or not. Accordingly, the remote control is possible from out of the radiation controlled area and the inspection work can be simplified. (Seki, T.)

Electrochemical noise measurements under pressurized water reactor conditions

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2000-01-01

Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

A CAMAC based real-time noise analysis system for nuclear reactors

International Nuclear Information System (INIS)

Ciftcioglu, O.

1987-01-01

A CAMAC based real-time noise analysis system was designed for the TRIGA MARK II nuclear reactor at the Institute for Nuclear Energy, Istanbul. The input analog signals obtained from the radiation detectors are introduced to the system through CAMAC interface. The signals coverted into digital form are processed by a PDP-11 computer. The fast data processing based on auto/cross power spectral density computations is carried out by means of assembly written FFT algorithms in real-time and the spectra obtained are displayed on a CAMAC driven display system as an additional monitoring device. The system has the advantage of being software programmable and controlled by a CAMAC system so that it is operated under porgram control for reactor surveillance, anomaly detection and diagnosis. The system can also be used for the identification of nonstationary operational characteristics of the reactor in long term by comparing the noise power spectra with the corresponding reference noise patterns prepared in advance. (orig.)

Developments and application of neutron noise diagnostics of sodium cooled fast reactors

International Nuclear Information System (INIS)

Zylbersztejn, F.

2013-01-01

The Sodium cooled Fast Reactor (SFR) is one of the six reactor types selected by the Generation-IV international forum (GIF), and the building of an industrial prototype is planned in France. The safety standard of the future SFR has to be equivalent to the EPR's. The general improvement of the safety of the new reactor goes through the examination of all the potentially harmful scenarios and both the study and monitoring of early signs. The mechanical deformations of the core can have harmful consequences in sodium fast reactors, such as unexpected power variations due to the reactivity increase in case of core compaction, or the excessive deterioration of the mechanical structures. The monitoring of such phenomena and of their potential early signs is then needed. The monitoring of such phenomena can be done with neutron detectors placed inside and outside the tank. This PhD thesis deals with the study of the neutron noise generated by the periodic deformation of the SFR core, restricted to the so-called core compaction or core flowering phenomenon, a deformation consisting in the variation of the inter-assembly sodium width by a radial bending the assemblies (the assemblies in SFR are held by the base). The PhD thesis has been performed within collaboration between CEA (France) and Chalmers Institute of Technology (Sweden). The work realized during the thesis led to the publication of 3 articles as first author and another as second author. This work has embraced the following topics: A state of the art of the monitoring of the core deformation phenomenon by interpretation of the noise measurements in SFR has been done. The PHENIX reactor multi physics measurements database has been scrutinized to provide an interpretation of the neutron noise bringing out mechanical vibration phenomena. An important conclusion was that the lack of theoretical knowledge about the neutron noise induced by the vibration phenomenon and the ill positioning of the neutron detectors

Noise thermometry in nuclear reactors

International Nuclear Information System (INIS)

Hoewener, H.

1985-08-01

Since in nuclear reactors the measuring sensor cannot be easily replaced, the value of the sensor resistance, as well as the selection of transmission lines with respect to good transmission characteristics of the whole arrangement and minimizing the correlative error terms, must already be optimized when designing a noise thermometer arrangement. The TRARAU computer program was developed for this purpose enabling the influences of the lines to be computed by taking into consideration all the effects occurring through the lines, such as transmission errors and correlative error terms. In order to check the accuracy of the TRARAU computer program a series of laboratory measurements were implemented enabling both the pure transmission behaviour of the line arrangement with respect to the measuring signal to be detected, as well as the overall line error. In all cases this resulted in a very good agreement of the measured values with the computed values. The transmission behaviour of noise thermometer arrangements occuring in practice were studied with the example of two reactor experiments. In both cases it was possible to demonstrate successfully the potential of the computer program TRARAU. As the parametric studies have shown, optimum matching over unlimited band widths is not feasible in principle. By reducing the upper band limit, however, the line error can practically always be kept sufficiently small. With good matching larger band widths can also be used. (orig./HP) [de

Summary of the fourth conference on United States utility experience in reactor noise analysis

International Nuclear Information System (INIS)

Fry, D.N.

1987-01-01

The fourth informal conference on United States utility experience in reactor noise analysis and loose-part monitoring was held at the Northeast Utilities Service Company offices in Hartford, Connecticut, May 12-14, 1987. Host and general chairman for the meeting was J.V. Persio of Northeast Utilities. This conference provided a forum where utilities could share information on reactor noise analysis on an informal basis. There were about 60 attendees at the meeting representing 10 utilities, 3 reactor vendors, 8 consulting organizations, and 4 universities and research laboratories. Twenty-three papers were presented at the conference, dealing with various aspects of loose-part monitoring, neutron noise analysis, and standards activities

18th informal meeting on reactor noise

International Nuclear Information System (INIS)

Dach, K.

1986-01-01

The proceedings contain the texts of 24 papers all of which fall under the INIS Subject Scope. The main purpose of the meeting of more than 90 scientists and specialists from 18 countries was for them to get acquainted with the current state and prospective development of noise diagnostics application in nuclear reactor operation. (Z.M.)

A study on the fault diagnostic techniques for reactor internal structures using neutron noise analysis

International Nuclear Information System (INIS)

Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk

1994-08-01

The unfavorable phenomena, such as flow induced vibration and aging process in reactor internals, cause degradation of structural integrity and may result in loosing some mechanical binding components which might impact other equipments and components or cause flow blockage. Since these malfunctions and potential failures change reactor noise signal, it is necessary to analyze reactor noise signal for early fault diagnosis in the point of few of safety and plant economics. The objectives of this study are to establish fault diagnostic and TS(thermal shield), and to develop a data acquisition and signal processing software system. In the first year of this study, an analysis technique for the reactor internal vibration using the reactor noise was proposed. With the technique proposed and the reactor noise signals (ex-core neutron and acceleration), the dynamic characteristics of Ulchin-1 reactor internals were obtained, and compared with those of Tricastin-1 which is the prototype of Ulchin-1. In the second year, a PC-based expert system for reactor internals fault diagnosis is developed, which included data acquisition, signal processing, feature extraction function, and represented diagnostic knowledge by the IF-THEN rule. To know the effect of the faults, the reactor internals of Ulchin-1 is modeled using FEM and simulated with an artificial defect given in the hold-down spring. Trend in the dynamic characteristics of reactor internals is also observed during one fuel cycle to know the effect of boron concentration. 100 figs, 7 tabs, 18 refs. (Author)

Status of sodium boiling noise detection programme at reactor research centre, India

Energy Technology Data Exchange (ETDEWEB)

Prabhakar, R; Elumalai, G [Reactor Engineering Laboratory, Reactor Research Centre, Chingleput, Tamil Nadu (India)

1982-01-01

Acoustic detection of sodium boiling is a promising technique to monitor subassembly fault in a last reactor. This paper summarises the programme for developing this detection system and describes the design of a high temperature transducer for boiling detection. It is appreciated that the background noise from primary pumps can interfere with this detection. Noise measurements were therefore carried out during water testing of the primary pump of the Fast Breeder Test Reactor. Some preliminary results of these measurements are presented.

Status of sodium boiling noise detection programme at reactor research centre, India

International Nuclear Information System (INIS)

Prabhakar, R.; Elumalai, G.

1982-01-01

Acoustic detection of sodium boiling is a promising technique to monitor subassembly fault in a last reactor. This paper summarises the programme for developing this detection system and describes the design of a high temperature transducer for boiling detection. It is appreciated that the background noise from primary pumps can interfere with this detection. Noise measurements were therefore carried out during water testing of the primary pump of the Fast Breeder Test Reactor. Some preliminary results of these measurements are presented

Measurement and analysis of the neutron noise of the pool research reactor at IPEN

International Nuclear Information System (INIS)

Simoes, Graciete Pedro

1979-01-01

Variations in the neutron density or power of a nuclear reactor (the neutron noise) operating at nominally constant power are generally random and can only be described in terms of statistical parameters. Random variations in the power of a power reactor are produced by one or more driving functions. In this work the neutron noise of the pool reactor IEAR-1 (2 MW nominal power) has been studied using two compensated ionization chambers ( Westinghouse VJL6377) and related to three possible-driving functions, namely vibration of the control bar and reactor support bridge and the temperature of the water entering the core. The CIC detectors were located in rigid tubes in turn positively located in the reactor lattice plate. Conventional accelerometers were used. Temperature measurements were made with a NiCr/Ni thermocouple (wire diam ∼ 0.2mm) located 10 mm above the top of a fuel element. Although the correlation between the measured neutron signals was high ( > 0,4) for frequencies in the range 0 to 10 Hz no resonances were identified in the neutron noise. A significant correlation (> 0,4) between the control bar acceleration and the neutron flux was obtained in the frequency range 0 to 10 Hz. The measured correlation between the neutron noise and both the bridge vibration and the reactor water inlet temperature was insignificant. (author)

Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; van Dam, H.; Kleiss, E.B.J.; van Uitert, G.C.; Veldhuis, D.

1982-01-01

The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations.

Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1982-01-01

The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations. (author)

Two-detector cross-correlation noise technique and its application in measuring reactor kinetic parameters

International Nuclear Information System (INIS)

Lu Guiping; Peng Feng; Yi Jieyi

1988-01-01

The two-detector cross-correlation noise technique is a new method of measuring reactor kinetic parameters developed in the sixties. It has the advantages of non-perturbation in core, high signal to noise ratio, low space dependent effect, and simple and reliable in measurement. A special set of cross-correlation analyzer has been prepared for measuring kinetic parameters of several reactor assemblies, such as the High Flux Engineering Test Reactor, its zero power mock up facility and a low enriched uranium light water lattice zero power facility

Application of noise analysis on the operation of the RSG-GAS reactor

International Nuclear Information System (INIS)

Surian P, Tukiran S

1999-01-01

The RSG-GAS reactor has been operating for radioisotope production and experiments so that it is necessary to perform analysis of the reactor operation. analysis was done based on reactor noise experiment. Neutron noise at low and high power of the RSG-GAS has been analyzed using time and frequency domain with aim to determine the safety of reactor operation. The safety of reactor operation based on two parameters as, prompt neutron decay constant and decay ratio. The parameters are useful for reactor operation, so it is necessary to determine accurately. For determining prompt neutron decay constant, neutron density in the core of reactor which operated at 10 k W, was collected by using Fission Chamber detectors (FC). Based on power spectral density (PSD) was achieved break frequency about 23 Hz, so that the prompt neutron decay constant is about 151 sec -1 . While at at high power 20 MW, neutron density was collected by using Compensated Ionization Chamber (CIC) detector. The result at high power showed that there is reactivity effect in the core because of fluctuation in temperature and density of the coolant, and the decay ratio of 0.20, showed that the reactor is still operation in stable

Development of electronic system for reactivity measurement and reactor noise analysis

International Nuclear Information System (INIS)

Strohl, Claude Emile; Soares, Adalberto Jose

1996-01-01

In nuclear power reactors, the neutron detector signal is dependent of the number of fissions and the reactor power level. The detector signal can be divided into two components; a DC component, proportional to the average value and an AC component, which is the fluctuating part superimposed to the DC component. The analysis of the fluctuating part of the signal is called noise analysis and allow us to investigate phenomena occurring within the reactor vessel, such as vibration of fuels elements and coolant density, temperature, pressure and flow changes. On the other hand, the measure of the static DC part allow us to measure the local power density. This paper describes the development of a personal computer based signal conditioning card that, together with a personal computer commercial data acquisition card, can be used for noise analysis measurements and reactivity measurements of signals coming from ionization chambers or SPD's. (author)

Multivariate statistical pattern recognition system for reactor noise analysis

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

1976-01-01

A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system

Multivariate statistical pattern recognition system for reactor noise analysis

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

1975-01-01

A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system. 19 references

Simulating the temperature noise in fast reactor fuel assemblies

International Nuclear Information System (INIS)

Kebadze, B.V.; Pykhtina, T.V.; Tarasko, M.Z.

1987-01-01

Characteristics of temperature noise at various modes of coolant flow in fast reactor fuel assemblies (FA) and for different points of sensor installation are investigated. Stationary mode of coolant flow and mode with a partial overlapping of FA through cross section, resulting in local temperature increase and sodium boiling, are considered. Numerical simulation permits to evaluate time characteristicsof temperature noise and to formulate requirements for dynamic characteristics of the sensors, and also to clarify the dependence of coolant distribution parameters on the sensor location and peculiarities of stationary temperature profile

SMORN-III benchmark test on reactor noise analysis methods

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Hirota, Jitsuya

1984-02-01

A computational benchmark test was performed in conjunction with the Third Specialists Meeting on Reactor Noise (SMORN-III) which was held in Tokyo, Japan in October 1981. This report summarizes the results of the test as well as the works made for preparation of the test. (author)

Investigations of the influence of feedback and coupling effects on neutron noise in a nuclear reactor

International Nuclear Information System (INIS)

Vath, W.

1975-01-01

Investigations are described of the influence of a known feedback, i.e., a control loop with known transfer function, on the spectra of neutron chamber signals. Theoretical formulas for the spectra are derived using the point reactor model. These formulas were verified by noise measurements in a zero power reactor. Special attention is given to the noise generated by the control loop, the influence of this feedback noise on the spectra being verified experimentally. In large reactors space dependent transfer functions must be taken into account. As a first approximation to handle the spatial dependence, the afore-mentioned investigations were extended to the two-point reactor model. Corresponding experimental work was done for the Argonaut Reactor Karlsruhe (ARK) with a symmetrical two-slab core loading. Coolant boiling in a BWR was investigated. The boiling must be considered as a feedback mechanism as well as an extrnal reactivity perturbation. In order to simulate the steam bubble content, nitrogen gas was injected into the water moderator of the ARK. By modulating the total gas flow according to the instantaneous reactor power the feedback effect was simulated. The gas flow produced a band-limited, white reactivity noise. The upper break frequency could be used to determine the travelling time of the bubbles through the core. (author)

A program for dynamic noise investigations of reactor systems

International Nuclear Information System (INIS)

Antonov, N.A.; Yaneva, N.B.

1980-01-01

A stochastic process analysis in nuclear reactors is used for the state diagnosis and dynamic characteristic investigation of the reactor system. A program DENSITY adapted and tested on an IBM 360 ES type computer is developed. The program is adjusted for fast processing of long series exploiting a relatively small memory. The testing procedure is discussed and the method of the periodic sequences corresponding to characteristic reactivity perturbations of the reactor systems is considered. The program is written for calculating the auto-power spectral density and the cross-power spectral density, as well as the coherence function of stationary statistical time series using the advantages of the fast Fourier transformation. In particular, it is shown that the multi-frequency binary sequences are very useful with respect to the signal-to-noise ratio and the frequency distribution in view of the frequency reactor test

Theory of fluctuations and parametric noise in a point nuclear reactor model

International Nuclear Information System (INIS)

Rodriguez, M.A.; San Miguel, M.; Sancho, J.M.

1984-01-01

We present a joint description of internal fluctuations and parametric noise in a point nuclear reactor model in which delayed neutrons and a detector are considered. We obtain kinetic equations for the first moments and define effective kinetic parameters which take into account the effect of parametric Gaussian white noise. We comment on the validity of Langevin approximations for this problem. We propose a general method to deal with weak but otherwise arbitrary non-white parametric noise. Exact kinetic equations are derived for Gaussian non-white noise. (author)

Noise analysis method for monitoring the moderator temperature coefficient of pressurized water reactors: Neural network calibration

International Nuclear Information System (INIS)

Thomas, J.R. Jr.; Adams, J.T.

1994-01-01

A neural network was trained with data for the frequency response function between in-core neutron noise and core-exit thermocouple noise in a pressurized water reactor, with the moderator temperature coefficient (MTC) as target. The trained network was subsequently used to predict the MTC at other points in the same fuel cycle. Results support use of the method for operating pressurized water reactors provided noise data can be accumulated for several fuel cycles to provide a training base

A report on the second specialists' meeting on reactor noise (SMORN II) Gatlinburg, Tennessee, September 19th-23rd 1977

International Nuclear Information System (INIS)

Greef, C.P.

1978-03-01

The current state of power reactor noise analysis as presented at the Second Specialists' Meeting on Reactor Noise (SMORN II) is reviewed. Practical successes include the analysis of PWR core support barrel motions and BWR instrument tube vibrations, leading to better understanding of the Physical processes involved. Future applications to LMFBRs are being actively pursued, but more information is required, notably on background noise sources in operating reactors. Theoretical understanding is improving in a number of specific areas, but general fundamental progress is hampered by the multiplicity of potential noise sources in a reactor. Past attempts to automate noise analysis so that simplified information may be presented to the reactor operator have not been totally successful. This has resulted from an over-hasty application of the techniques before all of the factors involved in physical phenomena were fully appreciated. (author)

On-line analysis of reactor noise using time-series analysis

International Nuclear Information System (INIS)

McGevna, V.G.

1981-10-01

A method to allow use of time series analysis for on-line noise analysis has been developed. On-line analysis of noise in nuclear power reactors has been limited primarily to spectral analysis and related frequency domain techniques. Time series analysis has many distinct advantages over spectral analysis in the automated processing of reactor noise. However, fitting an autoregressive-moving average (ARMA) model to time series data involves non-linear least squares estimation. Unless a high speed, general purpose computer is available, the calculations become too time consuming for on-line applications. To eliminate this problem, a special purpose algorithm was developed for fitting ARMA models. While it is based on a combination of steepest descent and Taylor series linearization, properties of the ARMA model are used so that the auto- and cross-correlation functions can be used to eliminate the need for estimating derivatives. The number of calculations, per iteration varies lineegardless of the mee 0.2% yield strength displayed anisotropy, with axial and circumferential values being greater than radial. For CF8-CPF8 and CF8M-CPF8M castings to meet current ASME Code S acid fuel cells

Application of noise analysis methods in nuclear reactor diagnostics

International Nuclear Information System (INIS)

Dach, K.

1985-01-01

By statistical evaluation of the fluctuation component of signals from selected detectors, noise diagnostics detects conditions of equipment which might later result in failure. The objective of early diagnostics is to detect the failed integrity of primary circuit components, failed detectors or anomalies of the thermohydraulic process. The commonest method of experimental data analysis is spectral analysis in the frequency range 0 to 50 Hz. Recently, expert diagnostic systems have been built based on artificial intelligence systems. Czechoslovakia participates in the experimental research of noise diagnostics in the context of the development of diagnostic assemblies for WWER-440 reactors. (M.D.)

Noise Diagnostics of Stationary and Non-Stationary Reactor Processes

Energy Technology Data Exchange (ETDEWEB)

Sunde, Carl

2007-04-15

This thesis concerns the application of noise diagnostics on different problems in the area of reactor physics involving both stationary and non-stationary core processes. Five different problems are treated, divided into three different parts. The first problem treated in the first part is the classification of two-phase flow regimes from neutron radiographic and visible light images with a neuro-wavelet algorithm. The algorithm consists of wavelet pre-processing and of an artificial neural network. The result indicates that the wavelet pre-processing is improving the training of the neural network. Next, detector tubes which are suspected of impacting on nearby fuel-assemblies in a boiling water reactor (BWR) are identified by both a classical spectral method and wavelet-based methods. It was found that there is good agreement between the different methods as well as with visual inspections of detector tube and fuel assembly damage made during the outage at the plant. The third problem addresses the determination of the decay ratio of a BWR from the auto-correlation function (ACF). Here wavelets are used, with some success, both for de-trending and de-nosing of the ACF and also for direct estimation of the decay ratio from the ACF. The second part deals with the analysis of beam-mode and shell-mode core-barrel vibrations in pressurised water reactors (PWRs). The beam-mode vibrations are analysed by using parameters of the vibration peaks, in spectra from ex core detectors. A trend analysis of the peak amplitude shows that the peak amplitude is changing during the fuel cycle. When it comes to the analysis of the shell-mode vibration, 1-D analytical and numerical calculations are performed in order to calculate the neutron noise induced in the core. The two calculations are in agreement and show that a large local noise component is present in the core which could be used to classify the shell-mode vibrations. However, a measurement made in the PWR Ringhals-3 shows

Noise Diagnostics of Stationary and Non-Stationary Reactor Processes

International Nuclear Information System (INIS)

Sunde, Carl

2007-01-01

This thesis concerns the application of noise diagnostics on different problems in the area of reactor physics involving both stationary and non-stationary core processes. Five different problems are treated, divided into three different parts. The first problem treated in the first part is the classification of two-phase flow regimes from neutron radiographic and visible light images with a neuro-wavelet algorithm. The algorithm consists of wavelet pre-processing and of an artificial neural network. The result indicates that the wavelet pre-processing is improving the training of the neural network. Next, detector tubes which are suspected of impacting on nearby fuel-assemblies in a boiling water reactor (BWR) are identified by both a classical spectral method and wavelet-based methods. It was found that there is good agreement between the different methods as well as with visual inspections of detector tube and fuel assembly damage made during the outage at the plant. The third problem addresses the determination of the decay ratio of a BWR from the auto-correlation function (ACF). Here wavelets are used, with some success, both for de-trending and de-nosing of the ACF and also for direct estimation of the decay ratio from the ACF. The second part deals with the analysis of beam-mode and shell-mode core-barrel vibrations in pressurised water reactors (PWRs). The beam-mode vibrations are analysed by using parameters of the vibration peaks, in spectra from ex core detectors. A trend analysis of the peak amplitude shows that the peak amplitude is changing during the fuel cycle. When it comes to the analysis of the shell-mode vibration, 1-D analytical and numerical calculations are performed in order to calculate the neutron noise induced in the core. The two calculations are in agreement and show that a large local noise component is present in the core which could be used to classify the shell-mode vibrations. However, a measurement made in the PWR Ringhals-3 shows

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

Investigations of SPND noise signals in VVER-440 reactors

International Nuclear Information System (INIS)

Kiss, S.; Lipcsei, S.; Hazi, G.

2001-01-01

This paper describes and characterises SPND noise measurements of an operating VVER-440 nuclear reactor. Characteristics of the signal can be radically influenced by the geometrical properties of the detector and the cable and by the measuring arrangement. Structure of phase spectra showing propagating perturbations measured on uncompensated SPN detectors is studied through models.(author)

Reactor internals vibration monitoring by neutron noise methods in PWRs

International Nuclear Information System (INIS)

Pazsit, I.; Por, G.; Lux, I.

1983-01-01

Certain elements of PWR cores such as control/fuel rods or cassettes, or other parts of reactor internals, often represent a vibration problem. Early analyses at operating PWR plant revealed that these vibrations can be detected by in-core neutron detectors, opening up the possibility of vibration monitoring and diagnostics by noise methods. Theoretical methods of calculating vibration induced neutron noise and its application to vibration diagnostics are summarized. Experiments to check theoretical conclusions are under way at the Central Research Institute for Physics, Budapest. (author)

Research program in reactor core diagnostics with neutron noise methods: Stage 3. Final report

International Nuclear Information System (INIS)

Pazsit, I.; Garis, N.S.; Karlsson, J.; Racz, A.

1997-09-01

Stage 3 of the program has been executed 96-04-12. The long term goal is to develop noise methods for identification and localization of perturbations in reactor cores. The main parts of the program consist of modelling the noise source, calculation of the space- and frequency dependent transfer function, calculation of the neutron noise via a convolution of the transfer function of the system and the noise source, i.e. the perturbation, and finally finding an inversion or unfolding procedure to determine noise source parameters from the neutron noise. Most previous work is based on very simple (analytical) reactor models for the calculation of the transfer function as well as analytical unfolding methods. The purpose of this project is to calculate the transfer function in a more realistic model as well as elaborating powerful inversion methods that do not require analytical transfer functions. The work in stage 3 is described under the following headlines: Further investigation of simplified models for the calculation of the neutron noise; Further investigation of methods based on neural networks; Further investigation of methods for detecting the vibrations and impacting of detectors; Application of static codes for determination of the neutron noise using the adiabatic approximation

Current development in data acquision and processing system for reactor noise analysis in PUSPATI

International Nuclear Information System (INIS)

Mohamad Amin Sharifuldin Salleh.

1986-11-01

A data acquisition and processing system for reactor noise analysis is described. It consists of four-channel isolation amplifier, a seven-channel DC amplifier, a four-channel analog to digital converter, analog filters, a microcomputer system and a plotter. This system is being applied to investigate the reactor dynamics of the PUSPATI TRIGA MK II reactor. (author)

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

Science.gov (United States)

Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

2010-10-01

The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

Reactor noise analysis by statistical pattern recognition methods

International Nuclear Information System (INIS)

Howington, L.C.; Gonzalez, R.C.

1976-01-01

A multivariate statistical pattern recognition system for reactor noise analysis is presented. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, updating, and data compacting capabilities. System design emphasizes control of the false-alarm rate. Its abilities to learn normal patterns, to recognize deviations from these patterns, and to reduce the dimensionality of data with minimum error were evaluated by experiments at the Oak Ridge National Laboratory (ORNL) High-Flux Isotope Reactor. Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the pattern recognition system

Calculation of the neutron noise induced by periodic deformations of a large sodium-cooled fast reactor core

International Nuclear Information System (INIS)

Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.

2014-01-01

The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)

Characters of neutron noise in full-size molten salt reactor

International Nuclear Information System (INIS)

Wang, Jiangmeng; Cao, Xinrong

2015-01-01

Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

A comparative study of time series modeling methods for reactor noise analysis

International Nuclear Information System (INIS)

Kitamura, Masaharu; Shigeno, Kei; Sugiyama, Kazusuke

1978-01-01

Two modeling algorithms were developed to study at-power reactor noise as a multi-input, multi-output process. A class of linear, discrete time description named autoregressive-moving average model was used as a compact mathematical expression of the objective process. One of the model estimation (modeling) algorithms is based on the theory of Kalman filtering, and the other on a conjugate gradient method. By introducing some modifications in the formulation of the problem, realization of the practically usable algorithms was made feasible. Through the testing with several simulation models, reliability and effectiveness of these algorithms were confirmed. By applying these algorithms to experimental data obtained from a nuclear power plant, interesting knowledge about the at-power reactor noise was found out. (author)

Vibration and acoustic noise emitted by dry-type air-core reactors under PWM voltage excitation

Science.gov (United States)

Li, Jingsong; Wang, Shanming; Hong, Jianfeng; Yang, Zhanlu; Jiang, Shengqian; Xia, Shichong

2018-05-01

According to coupling way between the magnetic field and the structural order, structure mode is discussed by engaging finite element (FE) method and both natural frequency and modal shape for a dry-type air-core reactor (DAR) are obtained in this paper. On the basis of harmonic response analysis, electromagnetic force under PWM (Pulse Width Modulation) voltage excitation is mapped with the structure mesh, the vibration spectrum is gained and the consequences represents that the whole structure vibration predominates in the radial direction, with less axial vibration. Referring to the test standard of reactor noise, the rules of emitted noise of the DAR are measured and analyzed at chosen switching frequency matches the sample resonant frequency and the methods of active vibration and noise reduction are put forward. Finally, the low acoustic noise emission of a prototype DAR is verified by measurement.

The interpretation of neutron noise in boiling water reactors

International Nuclear Information System (INIS)

John, T.M.; Singh, O.P.

1985-01-01

Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

Investigation of neural network paradigms for the development of automatic noise diagnostic/reactor surveillance systems

International Nuclear Information System (INIS)

Korsah, K.; Uhrig, R.E.

1991-01-01

The use of artificial intelligence (AI) techniques as an aid in the maintenance and operation of nuclear power plant systems has been recognized for the past several years, and several applications using expert systems technology currently exist. The authors investigated the backpropagation paradigm for the recognition of neutron noise power spectral density (PSD) signatures as a possible alternative to current methods based on statistical techniques. The goal is to advance the state of the art in the application of noise analysis techniques to monitor nuclear reactor internals. Continuous surveillance of reactor systems for structural degradation can be quite cost-effective because (1) the loss of mechanical integrity of the reactor internal components can be detected at an early stage before severe damage occurs, (2) unnecessary periodic maintenance can be avoided, (3) plant downtime can be reduced to a minimum, (4) a high level of plant safety can be maintained, and (5) it can be used to help justify the extension of a plant's operating license. The initial objectives were to use neutron noise PSD data from a pressurized water reactor, acquired over a period of ∼2 years by the Oak Ridge National Laboratory (ORNL) Power Spectral Density RECognition (PSDREC) system to develop networks that can (1) differentiate between normal neutron spectral data and anomalous spectral data (e.g., malfunctioning instrumentation); and (2) detect significant shifts in the positions of spectral resonances while reducing the effect of small, random shifts (in neutron noise analysis, shifts in the resonance(s) present in a neutron PSD spectrum are the primary means for diagnosing degradation of reactor internals). 11 refs, 8 figs

Application of noise analysis technique for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors

International Nuclear Information System (INIS)

Shieh, D.J.; Upadhyaya, B.R.; Sweeney, F.J.

1987-01-01

A new technique, based on the noise analysis of neutron detector and core-exit coolant temperature signals, is developed for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors (PWRs). A detailed multinodal model is developed and evaluated for the reactor core subsystem of the loss-of-fluid test (LOFT) reactor. This model is used to study the effect of changing the sign of the moderator temperature coefficient of reactivity on the low-frequency phase angle relationship between the neutron detector and the core-exit temperature noise signals. Results show that the phase angle near zero frequency approaches - 180 deg for negative coefficients and 0 deg for positive coefficients when the perturbation source for the noise signals is core coolant flow, inlet coolant temperature, or random heat transfer

Development of an electrometer/amplifier and filter set for analysis of reactor noise

International Nuclear Information System (INIS)

Strohl, Claude Emile

1996-01-01

In nuclear power reactors, the neutron detector signal is dependent on the number of fissions and the reactor power level. The detector signal can be divided into two components: a D C component, proportional to the average value and an A C component, which is the fluctuating part superimposed to the D C component. The analysis of the fluctuating part of the signal is called noise analysis and allow us to investigate phenomena occurring within the reactor vessel, such as vibrational of fuel elements and coolant density, temperature, pressure and flow changes. On the other hand, the measure of the static D C part allows us to measure the local power density. This work describes the development of a personal computer based signal conditioning card that, together with a personal computer commercial data acquisition card, can be used for noise analysis and reactivity measurements of signals coming from ionization chambers or SPD's. (author)

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

International Nuclear Information System (INIS)

Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit

2009-12-01

This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

Energy Technology Data Exchange (ETDEWEB)

Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit (Chalmers Univ. of Technology, Dept. of Nuclear Engineering, SE-412 96 Goeteborg (Sweden))

2009-12-15

This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

Neutron noise in nuclear reactors; Le bruit neutronique des reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Blaquiere, A. [Institut National des Sciences et Techniques Nucleaires (France); Pachowska, R. [Universite Technique de Varsovie (Poland)

1961-06-15

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [French] La puissance d'un reacteur nucleaire, dans les conditions du regime, est affectee de fluctuations dont les causes sont tres diverses. Ce comportement aleatoire rentre dans le cadre general de l'etude des 'bruits'. Entre autres sources ce bruit, nous analysons ici les fluctuations dues: a) a l'emission discontinue des neutrons provenant d'une source autonome; b) a la multiplication des neutrons au sein du reacteur. La methode que nous introduisons exploite les analogies entre les lois qui regissent un reacteur nucleaire au regime et certains systemes radioelectriques, en particulier les circuits a boucle de reaction. Le reacteur est caracterise par sa 'bande passante' et est decrit comme un systeme soumis a une succession d'impulsions aleatoires. Dans les conditions de fonctionnement non lineaires, l'effet du bruit neutronique est precise en utilisant une fonctionnelle non lineaire, ce qui relie cette theorie a

Measurement of nuclear reactor noise at low power; Merenje nuklearnog reaktorskog suma na malim snagama

Energy Technology Data Exchange (ETDEWEB)

Velickovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1968-07-01

Theoretical interpretation of reactor noise experiments is based on stochastic model developed and described in this paper. Ratio l/{beta} as well as subcriticality level can be determined bu measuring transfer function. In this paper the ratio l/{beta} was determined directly from auto-correlation functions for different critical configurations of the RB zero power reactor core and not by transfer function. This simplified the procedure significantly. It was found that the 0.5 W power level is most suitable for experimental study of neutron fluctuations. In this case fluctuations are intense compared to noise of the detector and electronic devices used.

Analysis and development of deterministic and stochastic neutron noise computing techniques with applications to thermal and fast reactors

International Nuclear Information System (INIS)

Rouchon, Amelie

2016-01-01

Neutron noise analysis addresses the description of small time-dependent flux fluctuations induced by small global or local perturbations of the macroscopic cross-sections. These fluctuations may occur in nuclear reactors due to density fluctuations of the coolant, to vibrations of fuel elements, control rods, or any other structures in the core. In power reactors, ex-core and in-core detectors can be used to monitor neutron noise with the aim of detecting possible anomalies and taking the necessary measures for continuous safe power production. The objective of this thesis is to develop techniques for neutron noise analysis and especially to implement a neutron noise solver in the deterministic transport code APOLLO3 developed at CEA. A new Monte Carlo algorithm that solves the transport equations for the neutron noise has been also developed. In addition, a new vibration model has been developed. Moreover, a method based on the determination of a new steady state has been proposed for the linear and the nonlinear full theory so as to improve the traditional neutron noise theory. In order to test these new developments we have performed neutron noise simulations in one-dimensional systems and in a large pressurized water reactor with heavy baffle in two and three dimensions with APOLLO3 in diffusion and transport theories. (author) [fr

BWR noise spectra and application of noise analysis to FBR

International Nuclear Information System (INIS)

Nomura, T.

1975-01-01

Work related to noise analysis, in Tokyo Shibaura Electric Co. Ltd. (Toshiba) and Nippon Atomic Industry Group Co. Ltd. (NAIG) for the past several years is reviewed. After considering the Japan-United States Seminar on Reactor Noise Analysis in 1968, other subjects discussed were boiling water reactor noise analysis and work in relation to FBR. Parts of these are related to each other. For example, boiling detection and temperature fluctuations are problems pertinent to both fields. As the main problems in zero-power-reactor noise are now basically understood, only a brief description of the experiments involving the advanced two detector method is made. Focus is rather placed on the area of power plant noise. (author)

Measures of the zero power nuclear reactor's kinetic parameters with application of noise analysis

International Nuclear Information System (INIS)

Martins, F.R.

1992-01-01

The purpose of this work was to establish an experimental technique based on noise analysis for measuring the ratio of kinetic parameters β/ Λ and the power of the Zero Power Nuclear Reactor IPEN-MB 01. A through study of the microscopic and macroscopic noise analysis techniques has been carried out. The Langevin technique and the point kinetic model were chosen to describe the stochastic phenomena that occur in the zero power reactor. Measurements have been made using two compensated ionization chambers localized in the water reflector at symmetric positions in order to minimize spatial effects on the neutron flux fluctuation. Power calibrations based on the low frequency plateau of the cross-power spectral density has also been carried out. (author)

State of the art on reactor noise analysis

International Nuclear Information System (INIS)

Bernard, P.; Fry, D.; Stegemann, D.; Van Dam, H.

1986-01-01

This report is the result of the work of a task force sponsored by the NEA Committee on Reactor Physics (NEACRP) and is divided into six chapters: 1. Loose-Parts Detection and Acoustic Monitoring, 2. Thermal Hydraulics Surveillance, 3. Flow Measurements, 4. Vibration Monitoring, 5. Surveillance Systems and Evaluation Methods, and 6. System Dynamic Analysis. Each chapter summarizes the current situation in noise analysis techniques with emphasis on the following aspects: . physical quantities considered, . possible anomalies involved, . sensors used for the detection, and . conditions of applications. The remainder of each chapter discusses future trends and recommendations

Measurement of nuclear reactor noise at low power levels; Merenje nuklearnog reaktorskog suma na malim snagama

Energy Technology Data Exchange (ETDEWEB)

Velickovic, Lj; Petrovic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1968-11-15

Spatial and time dependence relation of neutrons from fission and other reactions in the reactor core result in non Poisson fluctuations of neutron density. Analytical formula developed for auto-correlation function includes physical parameters which characterize time behaviour of neutrons in the reactor system. Since auto-correlation function can be easily measured, it is a useful tool for experimental determination of these parameters. Noise from the ionization chamber was measured and analyzed by a digital computer. measurements were analyzed completely in the time domain (auto-correlation functions). This enabled separating the noise caused by neutron detection from the noise from neutron density fluctuation in the reactor. All the results can be analyzed by spatial independent reactor theory. Physical analysis of reactor noise was limited to determination of {beta}/l ratio from auto-correlation measurements at 0.5 W power level (RB reactor in Vinca). Three different reactor core lattices were analyzed (lattice pitch 8 cm, 11.3 cm and 14 cm). It was shown that parameter {beta}/l could be determined from auto-correlation measurements of neutron density with high precision (few percents) Prostorna i vremenska povezanost neutrona koji nastaju u fisiji i drugim procesima koji se desavaju u reaktoru dovodi do ne Poisson-ovih fluktuacija neutronske gustine. Analiticka formula, razvijena za autokorelacionu funkciju ovih fluktuacija, sadrzi fizicke parametre koji karakterisu vremensko ponasanje neutrona u reaktorskom sistemu. Kako autokorelaciona funkcija moze lako da se meri, ona je korisno sredstvo za eksperimentalno odredjivanje ovih parametara. Sum iz jonizacione komore digitalno je meren i analiziran u digitalnom racunaru. Merenja su kompletno analizirana u vremenskom domenu (autokorelacione funkeije). To je olaksalo razdvajanje suma izazvanog procesom neutronske detekcije od suma koji potice od fluktuacija neutronske gustine u reaktorskom sistemu. Svi rezultati

Nuclear reactor pump diagnostics via noise analysis/artificial neural networks

International Nuclear Information System (INIS)

Keyvan, S.; Rabelo, L.C.

1991-01-01

A feasibility study is performed on the utilization of artificial neural networks as a tool for reactor diagnostics. Reactor pump signals utilized in a wear-out monitoring system developed for early detection of degradation of pump shaft are analyzed as a semi-benchmark test to study the feasibility of neural networks for pattern recognition. The Adaptive Resonance Theory (ART 2) paradigm of artificial neural networks is applied in this study. The signals are collected signals as well as generated signals simulating the wear progress. The wear-out monitoring system applies noise analysis techniques, and is capable of distinguishing between these signals and providing a measure of the progress of the degradation. This paper presents the results of the analysis of these data via the ART 2 paradigm

Application of neural networks and neutron noise for diagnostics of reactor internals vibration

International Nuclear Information System (INIS)

Garis, N.S.; Pazsit, I.; Gloeckler, O.

1995-01-01

It has long been known that vibration of reactor internals, in particular excessive vibrations of control rods, can be detected via the neutron noise they induce. Noise measurements are actually suitable to determine important diagnostic parameters such as the location of the vibrating rod and the vibration amplitude. An algorithm was earlier elaborated for this purpose, which is based on inversion of the expression describing the neutron noise as a function of vibration parameters. This inversion procedure is nevertheless complicated and not always unique. It was investigated whether a properly trained neural network can perform the inversion more effectively. It was found that artificial neural networks can be trained effectively to perform vibration diagnostics from neutron noise data fast, effectively and reliably. The present paper gives a description of the development and use of the neural networks for purposes of vibration diagnostics

Reactor noise analysis based on nonlinear dynamic theory - application to power oscillation

International Nuclear Information System (INIS)

Suzudo, Tomoaki

1993-01-01

The information dimension is one of the simplest quantities that can be used to determine the asymptotic motion of the time evolution of a nonlinear system. The application of this quantity to reactor noise analysis is proposed, and the possibility of its application to power oscillation analysis is examined. The information dimension of this regime is equal to the number of independent oscillating modes, which is an intuitive physical variable. Time series data from computer experiments and experiments with an actual physical system are used for the analysis. The results indicate that the method is useful for a detailed analysis of reactor power oscillation

Joint Neutron Noise Measurements on Metallic Reactor Caliban

International Nuclear Information System (INIS)

Chapelle, Amaury; Authier, Nicolas; Pierre, Casoli; Richard, Benoit; Myers, Will; Hutchinson, Jesson; Sood, Avneet; Rooney, Brian

2013-06-01

The aim of the experiments concerning neutron noise measurements presented in this article is to compare the measured parameters to the simulated ones. The results of these measurements must therefore be very accurate, with controlled uncertainties. To determine the relative contribution of uncertainties to the final result, a table presents the prompt multiplication obtained by a French Team and a U.S. team. The different sources of uncertainties are then explored, distinguishing them between three categories, those linked to the experimental configuration, to the detection process and finally to the analysis process. These experiments improve the safety task of reactivity control far from criticality, with static methods, and the knowledge of the behaviour of a subcritical reactor. (authors)

Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Dykin, V. [Chalmers Univ. of Tech., Nuclear Engineering, Goteborg (Sweden)

2014-07-01

This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)

Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations

International Nuclear Information System (INIS)

Pazsit, I.; Dykin, V.

2014-01-01

This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods, Stage 18

International Nuclear Information System (INIS)

Pazsit, Imre; Nam, Tran Hoai; Dykin, Victor; Jonsson, Anders

2013-01-01

This report constitutes Stage 18 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The objective of the research program is to contribute to the strategic research goal of competence and research capacity by building up competence within the Department of Nuclear Engineering at Chalmers University of Technology, regarding reactor physics, reactor dynamics and noise diagnostics. The purpose is also to contribute to the research goal of giving a basis for SSM's supervision by developing methods for identification and localization of perturbations in reactor cores. Results up to Stage 17 were reported in SKI and SSM reports, as listed in the report's summary

Investigation of the local component of power-reactor noise via diffusion theory

International Nuclear Information System (INIS)

Kosaly, G.

1975-03-01

The aim of the paper is to provide a theoretical background for the phenomenological model, which postulates the existence of a local component in the neutron noise of a light water cooled boiling water reactor. After the introductory review of the phenomenological model, noise calculation are performed by help of the one-group and two-group diffusion theory. Only in the two-group diffusion model it is succeeded to find a term in the response to a propagating disturbance of density which results in a small volume of neutrons physical sensivity around the point of observation. The problem, whether this local component can be a dominating term in the solution or not, is investigated in the Appenix. (Sz.Z.)

Upgrading the reactor noise diagnostic systems at the Paks NPP

International Nuclear Information System (INIS)

Czibok, T.; Dezsoe, Z.; Kiss, K.; Krinizs, K.; Lipcsei, S.

2002-01-01

The paper reports on the actual step in upgrading process of the reactor noise diagnostic systems at Paks NPP. This step has mainly a technical character. Renewal of facilities for signal conditioning and for data acquisition is going on. Autonomous systems at each of the four reactor units will be able to acquire a set of data series which can be arbitrarily chosen from the whole set of several hundred in-core neutron and other signals. The autonomous systems can be remotely controlled by a central computer through the local network. Modularity and extensibility are important features of the new systems: the size of the set of available signals can be extended and new modules for more advanced evaluations can be installed later. Present plans for system hardware upgrading are outlined, together with some technical details of measurement control, data acquisition moduls and network communication.(abstract)

Development of a noise-based method for the determination of the moderator temperature coefficient of reactivity (MTC) in pressurized water reactors (PWRs)

International Nuclear Information System (INIS)

Demaziere, C.

2002-01-01

The Moderator Temperature Coefficient of reactivity (MTC) is an important safety parameter of Pressurized Water Reactors (PWRs). In most countries, the so-called at-power MTC has to be measured a few months before the reactor outage, in order to determine if the MTC will not become too negative. Usually, the at-power MTC is determined by inducing a change in the moderator temperature, which has to be compensated for by other means, such as a change in the boron concentration. An MTC measurement using the boron dilution method is analysed in this thesis. It is demonstrated that the uncertainty of such a measurement technique is so large, that the measured MTC could become more negative than what the Technical Specifications allow. Furthermore, this technique incurs a disturbance of the plant operation. For this reason, another technique relying on noise analysis was proposed a few years ago. In this technique, the MTC is inferred from the neutron noise measured inside the core and the moderator temperature noise measured at the core-exit, in the same or in a neighbouring fuel assembly. This technique does not require any perturbation of the reactor operation, but was nevertheless proven to underestimate the MTC by a factor of 2 to 5. In this thesis, it is shown, both theoretically and experimentally, that the reason of the MTC underestimation by noise analysis is the radially loosely coupled character of the moderator temperature noise throughout the core. A new MTC noise estimator, accounting for this radially non-homogeneous moderator temperature noise is proposed and demonstrated to give the correct MTC value. This new MTC noise estimator relies on the neutron noise measured in a single point of the reactor and the radially averaged moderator temperature noise measured inside the core. In the case of the Ringhals-2 PWR in Sweden, Gamma-Thermometers (GTs) offer such a possibility since in dynamic mode they measure the moderator temperature noise, whereas in static

Acoustic detection of a cavitation noise in the French breeder reactor Phenix

International Nuclear Information System (INIS)

Brunet, M.; Desprets, A.

1981-06-01

The French Phenix reactor is provided with an in-core multi-sensor acoustic surveillance system. But its efficiency with regard to early boiling detection is still to be proven. For lack of boiling events within the core, a cavitating dummy steel subassembly has been loaded into the reactor, as a simulation of boiling signal. Cavitation is controlled through a slow rise of the primary flow in various core conditions: isothermal situation and during a rise of power. Assuming that the signal to noise ratio is of the same order of magnitude for cavitation signals as for boiling ones, the response of several signal processing techniques is evaluated. Pulse connecting seems to be the most efficient conventional method while pattern recognition appears as a promising alternative solution

Current Mode Neutron Noise Measurements in the Zero Power Reactor CROCUS

Science.gov (United States)

Pakari, O.; Lamirand, V.; Perret, G.; Braun, L.; Frajtag, P.; Pautz, A.

2018-01-01

The present article is an overview of developments and results regarding neutron noise measurements in current mode at the CROCUS zero power facility. Neutron noise measurements offer a non-invasive method to determine kinetic reactor parameters such as the prompt decay constant at criticality α = βeff / λ, the effective delayed neutron fraction βeff, and the mean generation time λ for code validation efforts. At higher detection rates, i.e. above 2×104 cps in the used configuration at 0.1 W, the previously employed pulse charge amplification electronics with BF3 detectors yielded erroneous results due to dead time effects. Future experimental needs call for higher sensitivity in detectors, higher detection rates or higher reactor powers, and thus a generally more versatile measurement system. We, therefore, explored detectors operated with current mode acquisition electronics to accommodate the need. We approached the matter in two ways: 1) By using the two compensated 10B-coated ionization chambers available in CROCUS as operational monitors. The compensated current signal of these chambers was extracted from coremonitoring output channels. 2) By developing a new current mode amplification station to be used with other available detectors in core. Characteristics and first noise measurements of the new current system are presented. We implemented post-processing of the current signals from 1)and 2) with the APSD/CPSD method to determine α. At two critical states (0.5 and 1.5 W), using the 10B ionization chambers and their CPSD estimate, the prompt decay constant was measured after 1.5 hours to be α=(156.9 ± 4.3) s-1 (1σ). This result is within 1σ of statistical uncertainties of previous experiments and MCNPv5-1.6 predictions using the ENDF/B-7.1 library. The newsystem connected to a CFUL01 fission chamber using the APSDestimate at 100 mW after 33 min yielded α = (160.8 ± 6.3) s-1, also within 1σ agreement. The improvements to previous neutron noise

Diagnostic measurement on research reactors

International Nuclear Information System (INIS)

Dach, K.; Zbytovsky, A.

A comparison is made of noise experiments on zero power and power reactors. The general characteristics of noise experiments on power reactors is their ''passivity'', i.e., the experiment does not require any interruption of the normal operating regime of the reactor system. On zero power research reactors where the fission reaction constitutes the dominant noise source such conditions have to be created in the study of noise components as to make the investigated noise dominant and the noise of the fission reaction the background. The simultaneous use of both methods makes it possible to determine the spectral composition of reactivity fluctuations, which facilitates the identification of noise sources. The conditions are described of the recordability of noise components. The possibilities are listed provided for research work in Czechoslovakia and the possibility is studied of setting up an expert team to organize the respective experimental programme on an international scale. Power reactors manufactured in the GDR are considered as the suitable experimental base. (J.P.)

Research program in reactor core diagnostics with neutron noise methods: Stage 3. Final report; Forskningsprogram angaaende haerddiagnostik med neutronbrusmetoder. Etapp 3. Slutrapport

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Garis, N.S.; Karlsson, J.; Racz, A. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Reactor Physics

1997-09-01

Stage 3 of the program has been executed 96-04-12. The long term goal is to develop noise methods for identification and localization of perturbations in reactor cores. The main parts of the program consist of modelling the noise source, calculation of the space- and frequency dependent transfer function, calculation of the neutron noise via a convolution of the transfer function of the system and the noise source, i.e. the perturbation, and finally finding an inversion or unfolding procedure to determine noise source parameters from the neutron noise. Most previous work is based on very simple (analytical) reactor models for the calculation of the transfer function as well as analytical unfolding methods. The purpose of this project is to calculate the transfer function in a more realistic model as well as elaborating powerful inversion methods that do not require analytical transfer functions. The work in stage 3 is described under the following headlines: Further investigation of simplified models for the calculation of the neutron noise; Further investigation of methods based on neural networks; Further investigation of methods for detecting the vibrations and impacting of detectors; Application of static codes for determination of the neutron noise using the adiabatic approximation. 12 refs, 18 figs.

Acoustic noises of the BOR-60 reactor steam generators when simulating leaks with argon and steam

International Nuclear Information System (INIS)

Sokolov, V.M.; Golushko, V.V.; Afanas'ev, V.A.; Grebenkin, Yu.P.; Muralev, A.B.

1985-01-01

Background acoustic noises of stea generators in different operational regimes and noises of argon and steam small leads (about 0.1 g/s) are studied to determine the possibility of designing the acoustic system for leak detection in sodium-water steamgenerators. Investigations are carried out at the 30 MW micromodule steam generator being in operation at the BOR-60 reactor as well as at the 20 MW tank type steam generator. Immersed ransduceres made of lithium niobate 6 mm in-diameter and waveguide transducers made of a stainless steel in the form of rods 10 mm in-diameter and 500 mm long are used as acoustic monitors. It is shown that the leak noise is more wide-band than the background noise of the steam generator and both high and low frequencies appear in the spectrum. The use of monitors of different types results in similar conslusions inrelation to the character of background noises and leak signals (spectral density, signal to-noise ratio) in the ase of similar bandroidths of the transduceres. A conclusion is made that the change of operational regimes leads to changes of background noise level, which can be close to the reaction of

Detector point of view of reactor internal vibrations under Gaussian coloured random forces - the problem of fitting neutron noise experimental data

International Nuclear Information System (INIS)

Arnal, R.S.; Martin, G.V.; Gonzalez, J.L.M.-C.

1988-01-01

This paper studies the local vibrations of reactor components driven by Gaussian coloured and white forces, when nonlinear vibrations arise. We study also the important problem of noise sources, modelization and the noise propagation through the neutron field using the discrete ordinates transport theory. Finally, we study the effect of the neutron field upon the PSD (power spectral density) of the noise source and we analyse the problem of fitting neutron noise experimental data to perform pattern recognition analysis. (author)

Surveillance of a nuclear reactor by pattern recognition analysis of the neutronic noise. Experience on Phenix LMFBR

International Nuclear Information System (INIS)

Le Guillou, G.; Bernardin, B.

1980-05-01

The purpose of this paper is to present some results of pattern recognition methods applied to the problem of supervising the behaviour of a nuclear reactor, especially the sodium cooled fast breeder Phenix. The problem to solve can be divided into two parts: first, from a set of data about the behaviour of the reactor, we have to find consistent classes of functionning. These classes must be interpreted by physical considerations; then, we have to elaborate a simple classification algorithm, which can be used in-line, to improve safety systems of fast breeder reactors, by giving helpful information for decision making. The neutronic noise has been chosen for our study

Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

Energy Technology Data Exchange (ETDEWEB)

Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)

2013-05-06

The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

Neutron density fluctuations in point reactor systems with dichotomic reactivity noise

International Nuclear Information System (INIS)

Sako, Okitsugu

1984-01-01

The exactly solvable stochastic point reactor model systems are analyzed through the stochastic Liouville equation. Three kinds of model systems are treated: (1) linear system without delayed neutrons, (2) linear system with one-group of delayed neutrons, and (3) nonlinear system with direct power feedback. The exact expressions for the fluctuations of neutron density, such as the moments, autocorrelation function and power spectral density, are derived in the case where the colored reactivity noise is described by the dichotomic, or two state, Markov process with arbitrary correlation time and intensity, and the effects of the finite correlation time and intensity of the noise on the neutron density fluctuations are investigated. The influence of presence of delayed neutrons and the effect of nonlinearity of system on the neutron density fluctuations are also elucidated. When the reactivity correlation time is very short, the correlation time has almost no effect on the power spectral density, and the relative fluctuation of neutron density in the stationary state is not affected very much by the presence of delayed neutrons and also by the nonlinearity of system. On the other hand, if the reactivity correlation time is very long, the effect of the reactivity noise on the power spectral density appears at very low frequency, and the presence of delayed neutrons has an effect of reducing the neutron density fluctuations. (author)

Power measurement of the RA-3 reactor using the neutron noise technique and 16N

International Nuclear Information System (INIS)

Gomez, Angel

2003-01-01

This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the 16 N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the 16 N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

A study of power fluctuations in a flip fuel reactor using the technique of noise analysis

Energy Technology Data Exchange (ETDEWEB)

Randall, J D; Wood, G C; Edwards, M A [Texas A and M University (United States)

1974-07-01

The Nuclear Science Center Reactor at Texas A and M University has experienced minor power fluctuations when operating at or near 1 megawatt. A noise analysis system was developed to investigate these fluctuations assuming that void formation, primarily due to nucleate boiling, was the cause. Experiments were carried out to correlate boiling noise with power level, fission product poisoning, and pool temperature. Results show that void formation in the core is the probable cause of the fluctuations with the onset of boiling occurring at 400 Kw. Data was also obtained that indicated the presence of boiling in a standard TRIGA core. (author)

Reactor noise analysis applications and systems in WWER-440 and WWER-1000 type PWRs

International Nuclear Information System (INIS)

Por, G.

1998-01-01

This paper presents an introduction on different types of well selected noise diagnostic methods with their occurrence in WWER reactors with an analysis of their impact on operational safety and aging which affects the installations safety as well. The main objective is to attract the attention of NPP management staff dealing with safety, safety culture, maintenance, operation and quality assurance proving that such methods can give benefit not only to economy but impact safety of nuclear installations

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)

Simulation of reactor noise analysis measurement for light-water critical assembly TCA using MCNP-DSP

International Nuclear Information System (INIS)

Yamamoto, Toshihiro; Sakurai, Kiyoshi; Tonoike, Kotaro; Miyoshi, Yoshinori

2001-01-01

Reactor noise analysis methods using Monte Carlo technique have been proposed and developed in the field of nuclear criticality safety. The Monte Carlo simulation for noise analysis can be made by simulating physical phenomena in the course of neutron transport in a nuclear fuel as practically as possible. MCNP-DSP was developed by T. Valentine of ORNL for this purpose and it is a modified version of MCNP-4A. The authors applied this code to frequency analysis measurements performed in light-water critical assembly TCA. Prompt neutron generation times for critical and subcritical cores were measured by doing the frequency analysis of detector signals. The Monte Carlo simulations for these experiments were carried out using MCNP-DSP, and prompt neutron generation times were calculated. (author)

Utility experience in reactor noise analysis in German LWRs

International Nuclear Information System (INIS)

Schuette, A.; Weingarten, J.; Wach, D.

1988-01-01

Reactor noise analysis has a long history in Fed. Rep. of Germany. Since the early 60ies it was directed to early fault, malfunction and anomaly detection in nuclear power plants. When successful diagnoses showed the high potential and benefits including economic and safety aspects, also the utilities began to have a vivid concern in application of the developed methods and systems. The know-how transfer from specialized teams to the onsite personnel as well as the development of more ''intelligent'' online systems are important present and future tasks for further improvement of their efficiency. The diagnostic group of GRS is active in both fields and has close cooperation with several utilities to achieve these objectives. Modern signal processing equipment and access to a comprehensive knowledge base stored in data banks are provided by the data analysis center of GRS. The available capabilities are described and examples of analysis results within the cooperation utilities/GRS are given. (author)

Surveillance of instruments by noise analysis

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

Random fluctuations of neutron flux, temperature, and pressure in a reactor provide multifrequency excitation of the corresponding instrumentation chains. Mathematical descriptors suitable for characterizing the output, or noise, of the instrumentation are reviewed with a view toward using such noise in detecting instrument faults. Demonstrations of the feasibility of this approach in a number of reactors provide illustrative examples. Comparisons with traditional surveillance testing are made, and a number of advantages and some disadvantages of using noise analysis as a supplementary technique are pointed out

The Multi一physics Research on I ron一Core Vibration Noise of Power Reactor

Directory of Open Access Journals (Sweden)

LI U Ja

2017-02-01

Full Text Available On the basis of theoretical research releted to the magnetostriction and maxwell’.s equations，the fi- nite element coupling in the transient electromagnetic field coupling，structure and sound field coupling has been developed In thts paper by using the flnlte element sOftWare CO}IS01., Whleh establish a serles three-phase COT’e re- actor model, to analyzing the power frequency magnetic field distribution，core magnetostrictive displacement，max- well force displacement and sound pressure level of the three-phase series core reactor under the power frequency working state. According to transient magnetic field distribution in the simulation of the reactor，the magnetic flux density distribution inside the reactor and the vibration displacement distribution are calculated，the acoustic field distribution is measured alao. It is shown that physical field simulation results and measured data are basically in consisent by experiment，it is proved multi-physics coupling is an effective method for forecast of noise.

Applications of a surveillance and diagnostics methodology using neutron noise from a pressurized-water reactor

International Nuclear Information System (INIS)

Wood, R.T.; Miller, L.F.; Perez, R.B.

1992-01-01

Two applications of a noise diagnostic methodology were performed with ex-core neutron detector data from a pressurized-water reactor (PWR). A feedback dynamics model of the neutron power spectral denisty was derived from a low-order whole-plant physical model made stochastic with the Langevin technique. From a functional fit to plant data, the response of the dynamic system to changes in important physical parameters was evaluated by a direct sensitivity analysis. In addition, changes in monitored spectra were related to changes in physical parameters, and detection thresholds using common surveillance discriminants were determined. A resonance model was developed from perturbation theory to give the ex-core neutron detector response for small in-core mechanical motions in terms of a pole-strength factor, a resonance asymmetry (or skewness) factor, a vibration damping factor, and a frequency of vibration. The mechanical motion paramters for several resonances were determined by a functional fit of the model to plant data taken at various times during a fuel cycle and were tracked to determined trends that indicated vibrational changes of reactor internals. In addition, the resonance model gave the ability to separate the resonant components of the power spectral density after the parameters had been identified. As a result, the behavior of several vibration peaks was monitored over a fuel cycle. The noise diagnostic methodology illustrated by these applications can be used in monitoring the condition of the reactor system. Early detection of degraded mechanical components or undesirable operating conditions by using such surveillance and diagnostic techniques would enhance plant safety. 15 refs., 6 figs., 1 tab

Safety aspects of neutron noise diagnostics and loose parts monitoring in WWER reactors

International Nuclear Information System (INIS)

Por, G.

1997-01-01

In this paper our aim is to give very short introduction on different types of well selected noise diagnostics methods and then mentioning their occurrence in WWER reactors we analyze what impact they might have to operational safety and for ageing (which also affects on safety of the installations). We do not deny, that one of our main aim is to call the attention of management staff of NPP, which deals with safety, safety culture, maintenance and operation proving, that such methods and system can give not only benefit to economy but also impact on safety of nuclear installations. 14 refs, 6 figs

Implementation of the neutron noise technique for subcritical reactors using a new data acquisition system

International Nuclear Information System (INIS)

Bellino, Pablo A.; Gomez, Angel

2009-01-01

A new data acquisition system was designed and programmed for nuclear kinetics parameter estimations in subcritical reactors. The system allows using any of the neutron noise techniques, since it could store the whole information available in the neutron detection system. The α Rossi, α Feynman and spectral analysis methods were performed in order to estimate the prompt neutron decay constant (and hence the reactivity). The measurements were done in the nuclear research reactor RA-1, where introducing the control rods, different reactivity levels where reached (until -7 dollars). With the three methods used, agreement was found between the estimations and the reference reactivities in each level, even when the detector efficiency was low. All the measurements were performed with a high gamma flux, although the results were found to be satisfactory. (author)

Theoretical and experimental notes on noise phenomena of KUR

International Nuclear Information System (INIS)

Kishida, Kuniharu

1980-01-01

The classification of global or local noise is important in reactor noise analysis. The term of ''global'' or ''local'' corresponds to that of ''system size'' or ''cell size'' in statistical physics. On the other hand, point model or phase space description is used in time series analysis. If a time series model describing spatial behavior is established, it will serve to reactor diagnosis. The noise phenomena of KUR are discussed from these points of view. In other words, from experimental results, the point reactor picture is reasonable to neutronic aspect but quantitative problem remains in coolant temperature fluctuations. By taking into account a diffusion type model, the spatial dependence is discussed for the problem remaining in coolant temperature fluctuations. It is pointed out that the time-space picture is a crucial idea of reactor noise phenomena. (author)

Software development of the mechanical vibration monitoring system of the CNA I reactor internals by neutron noise technique

International Nuclear Information System (INIS)

Wentzeis, Luis M.; Calvo, Maria D.

2009-01-01

The neutron noise analysis technique is an important predictive maintenance tool for early detection of failures such as sensor malfunctions and incipient mechanical problems located in the reactor internals. This technique was applied successfully in Argentina since 1987. The FER-GAEN group dependent of the CNEA developed the measuring system to detect anomalies as early as possible. The magnitude of interest in this analysis is the fluctuating component of the neutron flux known as 'neutron noise'. In order to improve and facilitate the analysis, a new software code was developed for the data acquisition of the neutron noise signals and neutron spectra estimation in the frequency domain. The RMS values related with the internals vibrations are calculated from these spectra and are chronologically displayed, in order to detect any anomalous vibration or incipient detector malfunction as early as possible. (author)

Process for testing noise emission from containers or pipelines made of steel, particularly for nuclear reactor plants

International Nuclear Information System (INIS)

Votava, E.; Stipsits, G.; Sommer, R.

1982-01-01

In a process for noise emission testing of steel containers or pipelines, particularly for testing primary circuit components of nuclear reactor plants, measuring sensors and/or associated electronic amplifiers are used, which are tuned for receiving the frequency band of the sound emission spectrum above a limiting frequency f G , but are limited or non-resonant for frequency bands less than f G . (orig./HP) [de

Temperature noise characteristics of pressurized water reactors

International Nuclear Information System (INIS)

Sweeney, F.J.; Upadhyaya, B.R.

1984-01-01

The core exit temperature noise RMS is linearly related to the core ΔT at a commercial PWR and LOFT. Test loop observations indicate that this linear behavior becomes nonlinear with blockages, boiling, or power skews. The linear neutron flux to temperature noise phase behavior is indicative of a pure time delay process, which has been shown to be related to coolant flow velocity in the core. Therefore, temperature noise could provide a valuable diagnostic tool for the detection of coolant blockages, boiling, and sensor malfunction under both normal and accident conditions in a PWR

Identification and localization of absorbers of variable strength in nuclear reactors

International Nuclear Information System (INIS)

Demaziere, C.; Andhill, G.

2005-01-01

This paper investigates the possibility of localising a noise source of the type 'absorber of variable strength' (or reactor oscillator) from as few as five neutron detectors evenly distributed throughout the core of a commercial nuclear reactor. The novelty of this investigation lies with the fact that the calculations are performed for a realistic 2-D heterogeneous reactor in the 2-group diffusion approximation, via the prior determination of the corresponding reactor transfer function. It is first demonstrated that the response of such a reactor to a localized perturbation deviates significantly from point-kinetics. The space-dependence of the induced neutron noise thus carries enough information about the location of the noise source, which makes it possible to determine its position from a few detector readings. The identification of the type of noise source is easily performed from the in-phase behaviour of the induced neutron noise. Different unfolding techniques are finally tested. All these techniques rely on the use of the reactor transfer function. One of these techniques is based on the comparison between the actual measured neutron noise and the neutron noise calculated for every possible location of the noise source. This technique is very reliable and almost insensitive to the contamination of the detector signals by background noise, but also extremely CPU consuming. Another technique, based on the piece-wise inversion of the reactor transfer function and requiring little CPU effort, was developed. Although this technique is much less reliable when background noise is present, this technique is useful to indicate a region of the reactor where a noise source is likely to be located

Base neutron noise in PWRs

International Nuclear Information System (INIS)

Kosaly, G.; Albrecht, R.W.; Dailey, D.J.; Fry, D.N.

1981-01-01

Considerable activity has been devoted in recent years to the use of neutron noise for investigation of problems in pressurized-water reactors (PWRs). The investigators have found that neutron noise provides an effective way to monitor reactor internal vibrations such as vertical and lateral core motion; core support barrel and thermal shield shell modes, bending modes of fuel assemblies, and control rod vibrations. However, noise analysts have also concluded that diagnosis of a problem is easier if baseline data for normal plant operation is available. Therefore, the authors have obtained ex-core neutron noise signatures from eight PWRs to determine the similarity of signatures between plants and to build a base of data to determine the sources of neutron noise and thus the potential diagnostic information contained in the data. It is concluded that: (1) ex-core neutron noise contains information about the vibration of components in the pressure vessel; (2) baseline signature acquisition can aid understanding of plant specific vibration frequencies and provide a bases for diagnosis of future problems if they occur; and (3) abnormal core support barrel vibration can most likely be detected over and above the plant-to-plant signature variation observed thus far

Noise source analysis of nuclear ship Mutsu plant using multivariate autoregressive model

International Nuclear Information System (INIS)

Hayashi, K.; Shimazaki, J.; Shinohara, Y.

1996-01-01

The present study is concerned with the noise sources in N.S. Mutsu reactor plant. The noise experiments on the Mutsu plant were performed in order to investigate the plant dynamics and the effect of sea condition and and ship motion on the plant. The reactor noise signals as well as the ship motion signals were analyzed by a multivariable autoregressive (MAR) modeling method to clarify the noise sources in the reactor plant. It was confirmed from the analysis results that most of the plant variables were affected mainly by a horizontal component of the ship motion, that is the sway, through vibrations of the plant structures. Furthermore, the effect of ship motion on the reactor power was evaluated through the analysis of wave components extracted by a geometrical transform method. It was concluded that the amplitude of the reactor power oscillation was about 0.15% in normal sea condition, which was small enough for safe operation of the reactor plant. (authors)

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)

Relation between nonlinear or 'not-linear' characteristics in nuclear kinetics and noise analysis of neutron flux

International Nuclear Information System (INIS)

Kataoka, H.

1975-01-01

The 'not-linear' or '2nd-class-nonlinear' characteristics in nuclear reactor kinetics with the feedback effect in the high-power operation and induce the increase in the amplitude of the neutron flux noise, specially in the very low frequency region. The fundamental behaviour of 'not-linear' characteristics and its effect for the reactor noise was investigated. Application of the reactor noise analysis technique to power reactors has not been successful because of unknown large disagreement between the result of the conventional theoretical analysis and the experimental facts. When the cause of this discrepancy is clear, reactor noise analysis techniques can be effectively applied to instrumentation, control, monitoring and diagnosis of power reactors. (author)

Random processes in nuclear reactors

CERN Document Server

Williams, M M R

1974-01-01

Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

Replication of noise-sustained autocatalytic chemical structures.

Science.gov (United States)

Izús, Gonzalo G; Deza, Roberto R; Sánchez, Alejandro D

2010-06-21

Autocatalytic systems in a differential-flow reactor may undergo a differential-flow-induced chemical instability toward a convectively unstable regime, in which noise-sustained structures may appear. This is the case of a system with Gray-Scott kinetics in a packed-bed reactor, as reported in [B. von Haeften and G. Izus, Phys. Rev. E 67, 056207 (2003)]. In this work, two identical copies of such a system are coupled in master-slave configuration and submitted to independent spatiotemporal Gaussian white noise sources. Numerical simulation of two-dimensional reactors with uniform and Poiseuille flows reveals that the slave system replicates to a very high degree of precision and the convective patterns arising in the master one due to the presence of noise. The quality of this synchronization is assessed through several measures. A convective instability in the synchronization manifold is theoretically predicted and numerically confirmed.

Experiences in stability testing of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Otaduy, P.J.

1986-01-01

The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

Vibration and noise analysis in nuclear power plants

International Nuclear Information System (INIS)

1974-12-01

Results of the investigations on noise and vibration analysis are presented as a follow-up study of the work published in ''On-load Surveillance of Nuclear Power Plant Components by Noise and Vibration Analysis'' EUR 5036 e. The state of the art in on-load surveillance techniques of light water reactors is given by extending the preceding studies to investigations of boiling water reactors and by summarizing the latest results of pressurized water reactors, the basis being experimental and theoretical work performed by the different organizations involved in preparing this report. Finally, some developments with respect to measurement and identification methods are discussed

Model-based temperature noise monitoring methods for LMFBR core anomaly detection

International Nuclear Information System (INIS)

Tamaoki, Tetsuo; Sonoda, Yukio; Sato, Masuo; Takahashi, Ryoichi.

1994-01-01

Temperature noise, measured by thermocouples mounted at each core fuel subassembly, is considered to be the most useful signal for detecting and locating local cooling anomalies in an LMFBR core. However, the core outlet temperature noise contains background noise due to fluctuations in the operating parameters including reactor power. It is therefore necessary to reduce this background noise for highly sensitive anomaly detection by subtracting predictable components from the measured signal. In the present study, both a physical model and an autoregressive model were applied to noise data measured in the experimental fast reactor JOYO. The results indicate that the autoregressive model has a higher precision than the physical model in background noise prediction. Based on these results, an 'autoregressive model modification method' is proposed, in which a temporary autoregressive model is generated by interpolation or extrapolation of reference models identified under a small number of different operating conditions. The generated autoregressive model has shown sufficient precision over a wide range of reactor power in applications to artificial noise data produced by an LMFBR noise simulator even when the coolant flow rate was changed to keep a constant power-to-flow ratio. (author)

Dichotomous-noise-induced pattern formation in a reaction-diffusion system

Science.gov (United States)

Das, Debojyoti; Ray, Deb Shankar

2013-06-01

We consider a generic reaction-diffusion system in which one of the parameters is subjected to dichotomous noise by controlling the flow of one of the reacting species in a continuous-flow-stirred-tank reactor (CSTR) -membrane reactor. The linear stability analysis in an extended phase space is carried out by invoking Furutzu-Novikov procedure for exponentially correlated multiplicative noise to derive the instability condition in the plane of the noise parameters (correlation time and strength of the noise). We demonstrate that depending on the correlation time an optimal strength of noise governs the self-organization. Our theoretical analysis is corroborated by numerical simulations on pattern formation in a chlorine-dioxide-iodine-malonic acid reaction-diffusion system.

Fuel-assembly vibration-induced neutron noise in PWRs

International Nuclear Information System (INIS)

Sweeney, F.J.; Renier, J.P.

1983-01-01

Space-dependent reactor kinetics calculations were performed to interpret observed increases in the amplitude of pressurized water reactor (PWR), ex-core neutron detector noise with increasing fuel burnup and correspondingly decreasing soluble boron concentration. These noise amplitude increases have occurred at both low frequencies (< 1.0 Hz) and in the 2.0- to 4.0-Hz frequency range. The noise amplitude increases in the 2.0- to 4.0-Hz frequency range have usually been accompanied by a decrease in the fundamental mode fuel assembly resonant frequency from 3.5 to 2.5 Hz over a fuel cycle, which has also been attributed to grid spacer spring relaxation

Studies on multivariate autoregressive analysis using synthesized reactor noise-like data for optimal modelling

Energy Technology Data Exchange (ETDEWEB)

Ciftcioglu, O.; Hoogenboom, J.E.; Dam, H. van

1988-01-01

Studies on the multivariate autoregressive (MAR) analysis are carried out for the choice of the parameters for modelling the data obtained from various sensors optimally. Accordingly, the roles of the parameters on the analysis results are identified and the related ambiguities are reduced. Experimental investigations are carried out by means of synthesized reactor noise-like data obtained from a digital simulator providing simulated stochastic signals of an operating nuclear reactor so that the simulator constitutes a favourable tool for the present studies aimed. As the system is well defined with its known structure, precise comparison of the MAR analysis results with the true values is performed. With the help of the information gained through the studies carried out, conditions to be taken care of for optimal signal processing in MAR modelling are determined. Although the parameters involved are related among themselves and they have to be given different values suitable for the particular application in hand, some criteria, namely memory-time and sample length-time play an essential role in AR modelling and they are found to be applicable to each individual case commonly, for the establishment of the optimality.

Studies on multivariate autoregressive analysis using synthesized reactor noise-like data for optimal modelling

International Nuclear Information System (INIS)

Ciftcioglu, O.

1988-01-01

Studies on the multivariate autoregressive (MAR) analysis are carried out for the choice of the parameters for modelling the data obtained from various sensors optimally. Accordingly, the roles of the parameters on the analysis results are identified and the related ambiguities are reduced. Experimental investigations are carried out by means of synthesized reactor noise-like data obtained from a digital simulator providing simulated stochastic signals of an operating nuclear reactor so that the simulator constitutes a favourable tool for the present studies aimed. As the system is well defined with its known structure, precise comparison of the MAR analysis results with the true values is performed. With the help of the information gained through the studies carried out, conditions to be taken care of for optimal signal processing in MAR modelling are determined. Although the parameters involved are related among themselves and they have to be given different values suitable for the particular application in hand, some criteria, namely memory-time and sample length-time play an essential role in AR modelling and they are found to be applicable to each individual case commonly, for the establishment of the optimality. (author)

Effect of uncompensated SPN detector cables on neutron noise signals measured in VVER-440 reactors

Energy Technology Data Exchange (ETDEWEB)

Kiss, S. E-mail: kisss@sunserv.kfki.hu; Lipcsei, S. E-mail: lipcsei@sunserv.kfki.hu; Hazi, G. E-mail: gah@sunserv.kfki.hu

2003-03-01

The Self Powered Neutron Detector (SPND) noise measurements of an operating VVER-440 nuclear reactor are described and characterised. Signal characteristics may be radically influenced by the geometrical properties of the detector and the cable, and by the measuring arrangement. Simulator is used as a means of studying the structure of those phase spectra that show propagating perturbations measured on uncompensated SPN detectors. The paper presents measurements with detectors of very different sizes (i.e. 20 cm length SPNDs and the 200 cm length compensation cables), where the ratios of the global and local component differ significantly for the different detector sizes. This phenomenon is used up for signal compensation.

Identification of neutron noise sources in a boiling water reactor

International Nuclear Information System (INIS)

Sides, W.H. Jr.; Mathis, M.V.; Smith, C.M.

1977-01-01

Measurements were made at units 2 and 3 of the Browns Ferry Nuclear Power Plant in order to characterize the noise signatures of the neutron and process signals and to determine the usefulness of such signatures for anomaly detection in BWR-4s. Previous measurements and theoretical analyses of BWR noise by others were concerned with the determination of steam velocity and void fraction (using the local component of neutron noise) and with the sources of global noise. The work described is under a five-part program to develop a complete and systematic analysis and representation of BWR neutron and process noise through complementary measurements and stochastic model developments. The parts are: (1) recording as many neutron detector and process noise signals as are available in a BWR-4; (2) reducing these data to noise signatures in order to perform an empirical analysis of these signatures, and documenting the relationships between the signals from spatially separated neutron detectors and between neutron and process variables; (3) developing spatially dependent neutronic models coupled with thermal-hydraulic models to aid in interpreting the observed relationships among the measured noise signatures, (4) comparing measured noise signatures with model predictions to obtain additional insight into BWR-4 dynamic behavior and to validate the models; and (5) using these models to predict the sensitivity of noise monitoring for detection, surveillance, and diagnosis of postulated in-core anomalies in BWRs. The paper describes the procedures used to obtain the noise recordings and presents initial empirical analysis and observations pertaining to the noise signatures and the relationships between several noise variables in the 0.01- to 1-Hz range. The mathematical models have not been developed sufficiently to report theoretical results or to compare measured spectra with model predictions at this time

Modeling of the acoustic boiling noise of sodium during an assembly blockage in sodium-cooled reactors

International Nuclear Information System (INIS)

Vanderhaegen, M.

2013-01-01

In the framework of the fourth generation of nuclear reactors safety requirements, the acoustic boiling detection is studied to detect subassembly blockages. Boiling, that might occur during subassembly blockages and that can lead to clad failure, generates hydrodynamic noise that can be related to the two-phase flow. A bubble dynamics study shows that the sound source during subassembly boiling is condensation. This particular phenomenon generates most noise as a high subcooling is present in the subassembly and because of the high thermal diffusivity of sodium. This result leads to an estimate of the form of the acoustic spectrum that will be filtered and amplified during propagation inside the liquid. And even though it is unlikely that bubbles will be present inside the subassembly, due to the very gradual temperature profile at the wall and due to the geometry that leads to a strong confinement of the vapor, the historical bubble dynamics approach gives some insight in previous measurements. Additionally, some hypotheses can be disproved. These theoretical ideas are validated with a small water experiment, yet it also shows that a simple experience in sodium doesn't lead to a better knowledge of the acoustic source. A theoretical analysis also revealed that a realistic experiment with a simulant fluid, such as water or mercury, isn't representative. A similar conclusion is obtained when studying cavitation as a simulant acoustic source. As such, the acoustic detection of boiling, in comparison with other detection systems, isn't sufficiently developed yet to be applied as a reactor protective system. (author) [fr

Nuclear reactor kinetics and control

International Nuclear Information System (INIS)

Lewins, J.

1978-01-01

A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

Using noise diagnostics for detection and monitoring of vibrations of RPV internals, and for cause analysis in Eastern German NPP with WWER-440/230 type reactors of the Russian design type

International Nuclear Information System (INIS)

Schumann, P.

2000-10-01

The report explains in detail applications and results of noise diagnostic measurements for examining vibrations which caused damage to reactor pressure vessel internals. The vibrations investigated were observed in 1975 and 1985 at the WWER-440/230 type reactors of the Russian design installed in units 1 and 2 of the Greifswald nuclear power plant, and were identified as dominant control element vibrations and abnormal core basket movements. The report presents essential operating characteristics recorded during disturbed reactor operation, the inspection results, and characteristic experimental findings from noise signal analyses and special tests examining the physical aspects of the vibration processes and their causes. Some recommendations have been derived relating to monitoring and minimization of such vibrations. (orig./CB) [de

Neutron fluctuations in accelerator driven and power reactors via backward master equations

International Nuclear Information System (INIS)

Zhifeng Kuang

2000-05-01

The transport of neutrons in a reactor is a random process, and thus the number of neutrons in a reactor is a random variable. Fluctuations in the number of neutrons in a reactor can be divided into two categories, namely zero noise and power reactor noise. As the name indicates, they dominate (i.e. are observable) at different power levels. The reasons for their occurrences and utilization are also different. In addition, they are described via different mathematical tools, namely master equations and the Langevin equation, respectively. Zero noise carries information about some nuclear properties such as reactor reactivity. Hence methods such as Feynman- and Rossi-alpha methods have been established to determine the subcritical reactivity of a subcritical system. Such methods received a renewed interest recently with the advent of the so-called accelerator driven systems (ADS). Such systems, intended to be used either for energy production or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those of the traditionally used radioactive sources which were also assumed in the derivation of the Feynman- and Rossi-alpha formulae. Therefore it is necessary to re-derive the Feynman- and Rossi-alpha formulae. Such formulae for ADS have been derived recently but in simpler neutronic models. One subject of this thesis is the extension of such formulae to a more general case in which six groups of delayed neutron precursors are taken into account, and the full joint statistics of the prompt and all delayed groups is included. The involved complexity problems are solved with a combination of effective analytical techniques and symbolic algebra codes. Power reactor noise carries information about parametric perturbation of the system. Langevin technique has been used to extract such information. In such a treatment, zero noise has been neglected. This is a pragmatic

Vibration and noise diagnostics of nuclear power plants Pt. 1

International Nuclear Information System (INIS)

Nagy, I.

1991-01-01

Development, tasks and methodology of vibration and noise testing of nuclear power plants are overviewed. Reactor noise diagnostics methods are presented and their utilization at various reactors is summarized. Acoustic testing of primary circuit is also considered. Special attention is paid to leak detection and loose parts monitoring by acoustical testing methods. (R.P.) 11 refs.; 18 figs.; 1 tab

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 10. Final Report

International Nuclear Information System (INIS)

Demaziere, C.; Pazsit, I.; Sunde, S.; Wright, J.

2004-12-01

This report constitutes Stage 10 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The program executed in Stage 10 consists of three parts and the work performed in each part is summarized below. A so-called 2-D 2-group neutron noise simulator was previously developed at the Department of Reactor Physics. This simulator is able to calculate the space-dependence of the neutron noise induced by localised or spatially-distributed absorbers of variable strength or by vibrating absorbers. These calculations are performed in the 2-group diffusion approximation and can treat an arbitrary heterogeneous 2-D system. The goal of the present investigation is to use the simulator for unfolding purposes, i.e. to reconstruct the noise source from the detector readings. This task is particularly challenging since the number of detectors available in a commercial PWR can be very low. In this study, five detectors are assumed to be present and evenly distributed in the core. Furthermore, only localised absorbers of variable strength are considered as perturbations. Numerical simulations were first carried out to verify that the space-dependent local and global components of the neutron noise were overwhelmingly large compared to the point-kinetic term of the neutron noise whose spatial structure does not depend on the position of the perturbation. The significance of the space-dependent global component is that its relaxation length is large enough, so that several neutron detectors can monitor it. Therefore, the neutron noise induced at the position of the detectors is itself a function of the position of the perturbation. Unfolding procedures could thus be considered. Prior to performing any unfolding, the type of the noise source has to be determined, since the algorithms developed in this report are based on the assumption of an absorber of variable strength. Such a noise source can also be

Noise investigations in boiling water and pressurized water reactors

International Nuclear Information System (INIS)

Kosaly, G.

1979-08-01

The paper begins with the general discussion of the neutron noise measured by in-core and ex-core detectors. The theory of BWR noise is surveyed in detail. After the theoretical survey the physical background of the measuring methods applied in the fluctuation analysis of BWRs is discussed. The section on PWR noise discusses the information content of the neutron, pressure and displacement signals about the vibration of the vessel and its internals. (author)

Detection and localization of leak of pipelines of RBMK reactor. Methods of processing of acoustic noise

International Nuclear Information System (INIS)

Tcherkaschov, Y.M.; Strelkov, B.P.; Chimanski, S.B.; Lebedev, V.I.; Belyanin, L.A.

1997-01-01

For realization of leak detection of input pipelines and output pipelines of RBMK reactor the method, based on detection and control of acoustic leak signals, was designed. In this report the review of methods of processing and analysis of acoustic noise is submitted. These methods were included in the software of the leak detection system and are used for the decision of the following problems: leak detection by method of sound pressure level in conditions of powerful background noise and strong attenuation of a signal; detection of a small leak in early stage by high-sensitivity correlation method; determination of a point of a sound source in conditions of strong reflection of a signal by a correlation method and sound pressure method; evaluation of leak size by the analysis of a sound level and point of a sound source. The work of considered techniques is illustrated on an example of test results of a fragment of the leak detection system. This test was executed on a Leningrad NPP, operated at power levels of 460, 700, 890 and 1000 MWe. 16 figs

Simple formulae for interpretation of the dead time α (first moment) method of reactor noise

International Nuclear Information System (INIS)

Degweker, S.B.

1999-01-01

The Markov Chain approach for solving problems related to the presence of a non extending dead time in a particle counting circuit with time correlated pulses was developed in an earlier paper. The formalism was applied to, among others, the dead time α (first moment) method of reactor noise. For this problem, however the solution obtained was largely numerical in character and had a tendency to break down for systems close to criticality. In the present paper, simple analytical expressions are derived for the count rate and L ex , the quantities of interest in this method. Comparisons with Monte Carlo simulations show that these formulae are accurate in the range of system parameters of practical interest

Investigation on the applicability of Piety's on-line PSD-pattern recognition algorithm to boiling detection by neutron-noise at a swimming-pool reactor

International Nuclear Information System (INIS)

Behringer, K.; Spiekerman, G.; Yadigaroglu, G.

1984-11-01

The neutron noise signal of an initiation-of-boiling experiment performed at the SAPHIR reactor has been analyzed by the PSD-pattern recognition algorithm of Piety (1977); the results indicate that the onset of boiling can be detected by this method. Improved confidence statements for the statistical decision discriminants are given. (Auth.)

Determination of the decay constants and relative abundances of delayed neutrons by noise analysis in zero-power reactors

International Nuclear Information System (INIS)

Diniz, Ricardo

2005-01-01

A reactor noise approach has been employed at the IPEN/MB-01 research reactor facility in order to determine experimentally the effective delayed neutron parameters β i and λ i in a six group model and assuming the point reactor. The method can be considered a novice one because exploits the very low frequency domain of the spectral densities. The proposed method has some advantages to other in-pile methods since it does not disturb the reactor system and consequently does not 'excite' any sort of harmonic modes. As a byproduct and a consistency check, the β eff parameter was obtained without the need of the Diven factor and the power normalization and it is in excellent agreement with independent measurements. The theory/experiment comparison shows that for the abundances the JENDL 3.3 presents the best performance while for the decay constants the revised version of ENDF/B-VI.8 shows the best agreement. The best performance for the β eff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate β eff by as much as 4%. The β eff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235 U fission as made by Sakurai and Okajima. A new observed effect related to the correlation between the fluctuations of both measurement channels is also presented and discussed. This effect can be considered as an indirect evidence for the use of the point reactor model in this work as well as a possible useful tool in the understanding of reactor dynamics. (author)

Application of the neutron noise analysis technique in nuclear power plants

International Nuclear Information System (INIS)

Lescano, Victor H.; Wentzeis, Luis M.

1999-01-01

Using the neutron noise analysis in nuclear power plants, and without producing any perturbation in the normal operation of the plant, information of the vibration state of the reactor internals and the behavior of the operating conditions of the reactor primary circuit can be obtained. In Argentina, the neutron noise analysis technique is applied in customary way in the nuclear power plants Atucha I and Embalse. A database was constructed and vibration frequencies corresponding to different reactor internals were characterized. Reactor internals with particular mechanical vibrations have been detected and localized. In the framing of a cooperation project between Argentina and Germany, we participated in the measurements, analysis and modelisation, using the neutron noise technique, in the Obrigheim and Gundremmingen nuclear power plants. In the nuclear power plant Obrigheim (PWR, 350 M We), correlations between the signals measured from self-power neutron detectors and accelerometers located inside the reactor core, were made. In the nuclear power plant Gundremmingen (BWR, 1200 M We) we participated in the study of a particular mechanical vibration detected in one of the instrumentation tube. (author)

Power measurement of the RA-3 reactor using the neutron noise technique and {sup 16}N; Medicion de la potencia del reactor RA-3, mediante la tecnica de ruido neutronico y nitrogeno 16

Energy Technology Data Exchange (ETDEWEB)

Gomez, Angel [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. Reactores y Centrales Nucleares

2003-07-01

This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the {sup 16}N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the {sup 16}N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

First evaluation of low frequency noise measurements of in core detector signals in the measuring assembly Rheinsberg

International Nuclear Information System (INIS)

Collatz, S.

1982-01-01

Reactor noise spectra of in core neutron detectors are measured in the low frequency range (0.03 Hz to 1 Hz) and evaluated. The increase of the effective noise signal value is due to pressure oscillations or oscillations of special steam volume portions. Thus boiling monitoring of reactor cores in PWR type reactors may be possible, if the low frequency noise of the whole set of in core detectors is taken into account

Flow measurements using noise signals of axially displaced thermocouples

Energy Technology Data Exchange (ETDEWEB)

Kozma, R.; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

1990-01-01

Determination of the flow rate of the coolant in the cooling channels of nuclear reactors is an important aspect of core monitoring. It is usually impossible to measure the flow by flowmeters in the individual channels due to the lack of space and safety reasons. An alternative method is based on the analysis of noise signals of the available in-core detectors. In such a noise method, a transit time which characterises the propagation of thermohydraulic fluctuations (density or temperature fluctuations) in the coolant is determined from the correlation between the noise signals of axially displaced detectors. In this paper, the results of flow measurements using axially displaced thermocouples in the channel wall will be presented. The experiments have been performed in a simulated MRT-type fuel assembly located in the research reactor HOR of the Interfaculty Reactor Institute, Delft. It was found that the velocities obtained via temperature noise correlation methods are significantly larger than the area-averaged velocity in the single-phase coolant flow. Model calculations show that the observed phenomenon can be explained by effects due to the radial velocity distribution in the channel. (author).

Current applications of vibration monitoring and neutron noise analysis

International Nuclear Information System (INIS)

Damiano, B.; Kryter, R.C.

1990-02-01

Monitoring programs using vibration monitoring or neutron noise analysis have demonstrated the ability to detect and, in some cases, diagnose the nature of reactor vessel internals structural degradation. Detection of compromised mechanical integrity of reactor vessel internal components in its early stages allows corrective action to be taken before weakening or damage occurs. In addition to the economic benefits early detection and correction can provide, they can also help maintain plant safety. Information on the condition of reactor vessel internal components gained from a monitoring program supplements in-service inspection results and may be useful in justifying plant license extension. This report, which was prepared under the Nuclear Plant Aging Research Program sponsored by the US Nuclear Regulatory Commission, discusses the application of vibration monitoring and neutron noise analysis for monitoring light-water reactor vessel internals. The report begins by describing the effects of structural integrity loss on internals vibration and how measurable parameters can be used to detect and track the progress of degradation. This is followed by a description and comparison of vibration monitoring and neutron noise analysis, two methods for monitoring the mechanical integrity of reactor vessel internals condition monitoring programs in the United States, Federal Republic of Germany, and France, three countries having substantial commitments to nuclear power. The last section presents guidelines for US utilities wishing to establish reactor internals condition monitoring programs. 20 refs., 5 figs., 4 tabs

Multivariate distance: application to neutronic noise; Distancia multivariante: aplicacion al ruido neutronico

Energy Technology Data Exchange (ETDEWEB)

Blazquez, J.

2015-07-01

In a PWR type reactor, the neutron noise increases gradually its amplitude because of boron dilution. In a KWU type reactor, this phenomena is more intense and might cause to descend the operating power along the cycle. It is likely that this behavior will occur in the EPR prototype. Noise amplitude analysis is made in this work by using advanced statistics techniques, results are: criteria for the choice of the best sensor to be used in the Reactor Safety Logic; and early detection of sensor malfunctions. (Author)

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 8. Final report

International Nuclear Information System (INIS)

Pazsit, I.; Demaziere, C.; Arzhanov, V.

2003-01-01

This report constitutes Stage 8 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. A brief proposal for the continuation of this program in stage 9 is also given at the end of the report. The program executed in Stage 8 consists of three parts and the work performed in each part is summarized below. In Stage 7, an anomaly localisation algorithm was presented and applied to the Forsmark-1 channel instability event. This investigation was carried out with a set of static data that were typical of a generic BWR model, since no data except the Local Power Range Monitors (LPRMs) time signals recorded during the instability event were made accessible to us at that time. In the present Stage, these static data were obtained from Vattenfall Fuel AB, who used the SIMULATE-3 code. The first Section of this report is thus devoted to this anomaly localisation but this time with a relevant set of data, both regarding the static core conditions and regarding the LPRMs time signals. The noise source pointed out by the localisation algorithm was once again found to be very close to an unseated fuel assembly discovered during the fuel outage following the instability event. The localisation procedure relies in fact on the prior calculation of the reactor transfer function, carried out via the neutron noise simulator first presented in Stage 6 and later improved in Stage 7. The neutron noise simulator was also improved in this Stage. More precisely, a 2-D 2- group static core simulator was developed in order to calculate the static flux and the associated eigenvalue with a spatial discretization scheme compatible with the one in use in the neutron noise simulator, i.e. finite differences. It has generally been assumed that the Decay Ratio (DR), a parameter characterizing the stability of BWRs, is a space-independent parameter of the reactor, i.e. it is independent of which LPRM is used in the core to

Measurement of reactor parameters of the 'Nora' reactor by noise analysis method - power spectral density

International Nuclear Information System (INIS)

Jovanovic, S.; Stormark, E.

1966-01-01

Measurements of reactor parameters the Nora reactor by Power Spectral Density (PSD) method are described. In case of critical reactor this method was applied for direct measurement of β/l ratio, β is the effective yield of delayed neutrons and l is the neutron lifetime. In case of subcritical reactor values of α+β-ρ/l were measured, ρ is the negative reactivity. Out coming PSD was measured by a filter or by ISAC. PSD was registered by ISAC as well as the auto-correlation function [sr

Research and development program in reactor diagnostics and monitoring with neutron noise methods. Stage 6. Final report

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Demaziere, C.; Avdic, S.; Dahl, B. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Reactor Physics

2000-07-15

by a finite difference scheme, similar to the one used in the previous stage. Actually, a relatively large and involved computational tool was elaborated, containing several modules. In the first module the static, space-dependent group constants are generated, to be used as input by the finite difference dynamic code. These are calculated through CASMO/SIMULATE for a realistic heterogeneous core. The dynamic code then calculates the noise for a number of hypothetical perturbations. The perturbations can be given either as direct changes in the cross-sections, or changes in the temperature of various components (coolant, fuel etc.). In the latter case the temperature fluctuations are converted into cross-section fluctuations by another module. Finally the noise is calculated in two or three dimensions by using the static group constants, the time derivatives, the delayed neutron data and the perturbations data. By a suitable choice of the perturbation, the Green's function can also be calculated, from which the noise induced by other perturbations can be calculated by integration. The model was verified through comparison to analytical calculations. It has been proposed a long time ago that the cross-correlation or coherence between the neutron noise, measured locally in the core, and the core-exit thermocouples can be used to determine the MTC. Such a method would have large advantages over the traditional ones because it would not interfere with the reactor operation. Earlier investigations showed nevertheless that the MTC value, obtained from such investigations, gave a quantitative value for the MTC that was 2 to 5 times lower than the true one. One very likely reason for this deviation lies in the fact that in the evaluation of the noise measurement, it is assumed that the temperature fluctuations, driving the neutron noise, are homogeneously distributed in the core, and that the response of the reactor is point-kinetic. In reality none of these two

analysis of the environmental and socio-economic impact of noise

African Journals Online (AJOL)

Dr Obe

Interest in noise control studies is only recent in Nigeria. ... standards for environmental pollution by the Federal ... heaters, gas flare system, reactors, water injection plants, pumps ..... review of health effects of aircraft noise. Aust. NZ J. Public ...

Autonomous data acquisition system for Paks NPP process noise signals

International Nuclear Information System (INIS)

Lipcsei, S.; Kiss, S.; Czibok, T.; Dezso, Z.; Horvath, Cs.

2005-01-01

A prototype of a new concept noise diagnostics data acquisition system has been developed recently to renew the aged present system. This new system is capable of collecting the whole available noise signal set simultaneously. Signal plugging and data acquisition are performed by autonomous systems (installed at each reactor unit) that are controlled through the standard plant network from a central computer installed at a suitable location. Experts can use this central unit to process and archive data series downloaded from the reactor units. This central unit also provides selected noise diagnostics information for other departments. The paper describes the hardware and software architecture of the new system in detail, emphasising the potential benefits of the new approach. (author)

Audible Noise Measurement and Analysis of the Main Power Apparatus in UHV GIS Substations

Directory of Open Access Journals (Sweden)

Zhou Nian Guang

2016-01-01

Full Text Available Investigation of audible noise characteristics of the main power apparatus in UHV GIS substations provides essential statistics for the noise prediction and control. Noise pressure level, spectrum and attenuation characteristics of the main transformers and high voltage (HV reactors are measured and analyzed in this paper. The result shows that the main transformer and HV reactor have identical A-weighted equivalent sound pressure level. The medium- and low-frequency noises are the primary components in the spectral. More attention should be paid to the low-frequency bands in the noise control process. The noise of cooling fan has a large influence on that of the main transformer. Without the consideration of corona noise, the average A-weighted sound pressure level shows an overall decreasing trend with the increase of the propagation distance. Obvious interference phenomenon of the noises at 100 and 200Hz exists in the noise propagation process.

Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA1 and Embalse Argentine NPP

International Nuclear Information System (INIS)

Lescano, V.H.; Wentzeis, L.M.; Guevara, M.; Moreno, C.; Pineyro, J.

1996-01-01

In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

Study of the stochastic point reactor kinetic equation

International Nuclear Information System (INIS)

Gotoh, Yorio

1980-01-01

Diagrammatic technique is used to solve the stochastic point reactor kinetic equation. The method gives exact results which are derived from Fokker-Plank theory. A Green's function dressed with the clouds of noise is defined, which is a transfer function of point reactor with fluctuating reactivity. An integral equation for the correlation function of neutron power is derived using the following assumptions: 1) Green's funntion should be dressed with noise, 2) The ladder type diagrams only contributes to the correlation function. For a white noise and the one delayed neutron group approximation, the norm of the integral equation and the variance to mean-squared ratio are analytically obtained. (author)

Applications of noise analysis to nuclear safety

International Nuclear Information System (INIS)

Aguilar Martinez, Omar

2000-01-01

Noise Analysis techniques (analysis of the fluctuation of physical parameters) have been successfully applied to the operational vigilance of the technical equipment that plays a decisive role in the production cycle of a very complex industry. Although fluctuation measurements in nuclear installations started almost at the start of the nuclear era (see works by Feynman and Rossi on the development of neutron methodology), only recently have neutron noise diagnostic applications begun to be a part of the standard procedures for the performance of some modern nuclear installations. Following the relevant technical advances made in information sciences and analogical electronics, measuring the fluctuation of physical parameters has become a very effective tool for detecting, guarding and following up possible defects in a nuclear system. As the processing techniques for the fluctuation of a nuclear reactor's physical-neutron parameters have evolved (temporal and frequency analysis, multi-parameter self -regression analysis, etc.), the applications of the theory of non-lineal dynamics and chaos theory have progressed by focusing on the problem from another perspective. This work reports on those nuclear applications of noise analysis that increase nuclear safety in all types of nuclear facilities and that have been carried out by the author over the last decade, such as: -Void Force Critical Set Applications (Zero Power Reactor Applications, Central Institute of Physical Research, Budapest, Hungary); -Research Reactor Applications (Triga Mark III Reactor, National Institute of Nuclear Research, ININ, Mexico); -Power Reactor Applications in a Nuclear Power Plant (First Circuit of Block II, Paks Nuclear Center, Hungary); -Second Loop applications in a Nuclear Power Plant (Block I Paks Nuclear Center, Hungary; Block II Kalinin Nuclear Center, Russia); -Shield System Applications for the Transport of Radioisotopes (Nuclear Technology Center, Havana, Cuba) New trends in

Summary of the progress of reactor physics in Japan reviewing the activities related to NEA Committee on Reactor Physics

International Nuclear Information System (INIS)

Hirota, Jitsuya

1984-09-01

The progress of fast and thermal reactor physics, fusion neutronics and shielding researches in these twenty years can be clearly recognized in the reviews of reactor physics activities in Japan which had been perpared by the Special Committee on Reactor Physics: the joint committee under Atomic Energy Society of Japan and JAERI. Many topics of those discussed at the NEACRP meetings concerned fast reactor physics. Information exchange on the topics such as adjustment of group cross sections by integral data, central worth discrepancy, sodium void effect and heterogeneous core stimulated the researches in Japan. And achievements in Japan including those in the JAERI Fast Critical Facility FCA were reported and contributed largely to the international co-operation. In addition, the contribution from Japan was also made concerning a study of fusion blanket. Among various specialists' meetings recommended by NEACRP, those on nuclear data and benchmarks for reactor shielding were often held since 1973 and helpful to the progress of shielding researches in Japan. The Third Specialists' Meeting on Reactor Noise (SMORN-III) was held in Tokyo in 1981, indicating the recent progress in safety-related applications of reactor noise analysis. The NEACRP benchmark tests were quite useful to the progress of reactor physics in Japan, which included the benchmark calculations of BWR lattice cell, key parameters and burn-up characteristics of a large LMFBR, FBR and PWR shielding, and so on. It may be noted that the benchmark test on reactor noise analysis methods was successfully conducted by Japan in connection with SMORN-III. In addition, the co-operation was positively made to the compilation of light water lattice data, and the preparation of reviews on actinide production and burn-up, and blanket physics. (J.P.N.)

Temperature noise analysis and sodium boiling detection in the fuel failure mockup

International Nuclear Information System (INIS)

Sides, W.H. Jr.; Fry, D.N.; Leavell, W.H.; Mathis, M.V.; Saxe, R.F.

1976-01-01

Sodium temperature noise was measured at the exit of simulated, fast-reactor fuel subassemblies in the Fuel Failure Mockup (FFM) to determine the feasibility of using temperature noise monitors to detect flow blockages in fast reactors. Also, acoustic noise was measured to determine whether sodium boiling in the FFM could be detected acoustically and whether noncondensable gas entrained in the sodium coolant would affect the sensitivity of the acoustic noise detection system. Information from these studies would be applied to the design of safety systems for operating liquid-metal fast breeder reactors (LMFBRs). It was determined that the statistical properties of temperature noise are dependent on the shape of temperature profiles across the subassemblies, and that a blockage upstream of a thermocouple that increases the gradient of the profile near the blockage will also increase the temperature noise at the thermocouple. Amplitude probability analysis of temperature noise shows a skewed amplitude density function about the mean temperature that varies with the location of the thermocouple with respect to the blockage location. It was concluded that sodium boiling in the FFM could be detected acoustically. However, entrained noncondensable gas in the sodium coolant at void fractions greater than 0.4 percent attenuated the acoustic signals sufficiently that boiling was not detected. At a void fraction of 0.1 percent, boiling was indicated only by the two acoustic detectors closest to the boiling site

Systems for noise diagnostics of WWER nuclear power plants

International Nuclear Information System (INIS)

Por, G.

1996-01-01

The aim of this paper is to give a short overview of the noise diagnostics system developed by Hungarian firms which are in operation in WWER type NPP Units. Giving a list of systems developed for noise diagnostics of WWER reactors we present their main characteristics, their goal and some of their achievements. The second part deals with the problem of acceptance of noise system by NPP and regulations. (author). 24 refs

Systems for noise diagnostics of WWER nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Por, G [Technical Univ. of Budapest, Budapest (Hungary)

1997-12-31

The aim of this paper is to give a short overview of the noise diagnostics system developed by Hungarian firms which are in operation in WWER type NPP Units. Giving a list of systems developed for noise diagnostics of WWER reactors we present their main characteristics, their goal and some of their achievements. The second part deals with the problem of acceptance of noise system by NPP and regulations. (author). 24 refs.

Tendencies in operating power reactors

International Nuclear Information System (INIS)

Brinckmann, H.F.

1987-01-01

A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

Report of the first United States conference on utility experience with neutron noise analysis

International Nuclear Information System (INIS)

Fry, D.N.; Horne, G.P.; Mayo, C.W.

1984-01-01

An informal meeting was held in Washington, D.C. on April 3 and 4, 1984, to discuss the current state of the art and experiences with neutron noise analysis in US pressurized water reactors (PWRs). The meeting was attended by 33 persons representing 11 utilities and 3 PWR reactor vendors as well as consultants, universities, and research laboratories. Presentations at the meeting covered several applications of neutron noise for diagnosing such things as vibrations induced by baffle jetting, detection of mechanical degradation of thermal shield supports, and electrical degradation of nuclear instrumentation channels. Twenty-one responses were obtained from a questionnaire circulated to all participants requesting their viewpoints and experiences regarding neutron noise analysis. The meeting participants concluded that a working group on neutron noise analysis should be formed to (1) establish a baseline library of neutron noise data, (2) provide a forum for communicating experiences with neutron noise surveillance, and (3) develop good practices and quality assurance procedures for neutron noise measurement and interpretation

The failure diagnoses of nuclear reactor systems

International Nuclear Information System (INIS)

Sheng Huanxing.

1986-01-01

The earlier period failure diagnoses can raise the safety and efficiency of nuclear reactors. This paper first describes the process abnormality monitoring of core barrel vibration in PWR, inherent noise sources in BWR, sodium boiling in LMFBR and nuclear reactor stability. And then, describes the plant failure diagnoses of primary coolant pumps, loose parts in nuclear reactors, coolant leakage and relief valve location

Review of fast reactor activities in Italy, April 1978

International Nuclear Information System (INIS)

Pierantoni, F.

1978-01-01

In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments

Review of fast reactor activities in Italy, April 1978

Energy Technology Data Exchange (ETDEWEB)

Pierantoni, F [CNEN Fast Reactor Programme, Bologna (Italy)

1978-07-01

In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments.

Dynamics analysis of a boiling water reactor based on multivariable autoregressive modeling

International Nuclear Information System (INIS)

Oguma, Ritsuo; Matsubara, Kunihiko

1980-01-01

The establishment of the highly reliable mathematical model for the dynamic characteristics of a reactor is indispensable for the achievement of safe operation in reactor plants. The authors have tried to model the dynamic characteristics of a reactor based on the identification technique, taking the JPDR (Japan Power Demonstration Reactor) as the object, as one of the technical studies for diagnosing BWR anomaly, and employed the multivariable autoregressive modeling (MAR method) as one of the useful methods for forwarding the analysis. In this paper, the outline of the system analysis by MAR modeling is explained, and the identification experiments and their analysis results performed in the phase 4 of the power increase test of the JPDR are described. The authors evaluated the results of identification based on only reactor noises, making reference to the results of identification in the case of exciting the system by applying artificial irregular disturbance, in order to clarify the extent in which the modeling is possible by reactor noises only. However, some difficulties were encountered. The largest problem is the one concerning the separation and identification of the noise sources exciting the variables from the dynamic characteristics among the variables. If the effective technique can be obtained to this problem, the approach by the identification technique based on the probability model might be a powerful tool in the field of reactor noise analysis and the development of diagnosis technics. (Wakatsuki, Y.)

Undergraduate reactor control experiment

International Nuclear Information System (INIS)

Edwards, R.M.; Power, M.A.; Bryan, M.

1992-01-01

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Light water reactor safety research project

International Nuclear Information System (INIS)

Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

1980-07-01

The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

Measuring the background acoustic noise in the BN-600 steam generator

International Nuclear Information System (INIS)

Yugaj, V.S.; Zhukovets, V.N.; Ivannikov, V.I.; Vylomov, V.V.; Ryabinin, F.; Chernykh, P.G.; Flejsher, Yu.V.

1987-01-01

Acoustic noises in the lower chambers of evaporation and intermediate overheating moduli of the BN-600 reactor steam generator are measured. Bachground noises are registered in the whole range of frequencies studied, from 0.63 to 160 kHz. The comparison of noise spectra in evaporator and overheater has revealed a certain difference. However the general tendency is the reduction of the noise level at high frequencies > 8 kHz. The increase of the noise level at low steam content is observed only in a narrow of frequency range of 3-6 kHz

Noise thermometry - a new temperature measuring method

International Nuclear Information System (INIS)

Brixy, H.; Hecker, R.; Rittinghaus, K.F.

1975-01-01

The thermal Johnson-Niquist noise is the basis of noise thermometry. This temperature measuring method is, e.g., of interest insofar as the noise thermometer gives absolute values as a primary thermometer and is in principle extensively independent of environmental influences and material properties. The resistance values of the measuring probe are about 10 Ohm to a few kOhm. The demands of electronics are high, the self-noise of the measuring apparatus must be as small as possible; a comparative measuring method is advantageous. 1 to 2,500 K are given as a possible temperature range. An accuracy of 0.1% could be achieved in laboratory measurements. Temperature measurements to be used in operation in a few nuclear reactors are mentioned. (HP/LH) [de

Modeling and analysis of neutron noise from an ex-core detector at a pressurized water reactor

International Nuclear Information System (INIS)

Wood, R.T.; Perez, R.B.

1991-01-01

Two applications of a noise diagnostic methodology were performed using ex-core neutron detector data from a pressurized water reactor (PWR). A feedback dynamics model of the neutron power spectral density (PSD) was derived from a low-order whole-plant physical model made stochastic using the Langevin technique. From a functional fit to plant data, the response of the dynamic system to changes in important physical parameters was evaluated by a direct sensitivity analysis. In addition, changes in monitored spectra were related to changes in physical parameters and detection thresholds using common surveillance discriminants were determined. A resonance model was developed from perturbation theory to give the ex-core neutron detector response for small in-core mechanical motions in terms of a pole-strength factor, a resonance asymmetry (or skewness) factor, a vibration damping factor, and a frequency of vibration. The mechanical motion parameters for several resonances were determined by a functional fit of the model to plant data taken at various times during a fuel cycle and were tracked to determine trends that indicated vibrational changes of reactor internals. In addition, the resonance model gave the ability to separate the resonant components of the PSD after the parameters had been identified. As a result, the behavior of several vibration peaks were monitored over a fuel cycle. 9 refs., 6 figs., 1 tab

Lifetime measurement of prompt neutrons using the neutronic noise analysis

International Nuclear Information System (INIS)

Ortiz Servin, J.J.

1992-01-01

The purpose of this work is to estimate the life of the prompt neutrons, i, of a nuclear reactor utilizing the neutron noise analysis. This technique carry to development of mathematical model that is valid for lower powers reactor. The equation resulting convey to the observation about power spectrum behaviour respect to the frecquency. In this case, the reactor in study is the Triga Mark III of Nuclear Center of Mexico that it was provided of fission chambers for register the neutron fluxes. These fluxes was digitized and storage in computer disc as signals dependents of time, for later apply the Fourier Transformation and obtain the spectras. The spectras measured to different reactor powers were adjusted to the development equation before, using the method of square minimum and so estimate the parameter i. The analysis of results throw a value of 22.73 +/- 0.92 μs. On the other hand, the calculate value to the resolve the kinetic equation of reactor defer in lower than 4 % about the estimate. Of this, it concludes that the model utilized is trusty with a good mistake margin, moreover of that the technique of Neutron Noise analysis demonstrate be competitive (Author)

Modeling temperature noise in a fast-reactor pile

International Nuclear Information System (INIS)

Kebadze, B.V.; Pykhtina, T.V.; Tarasko, M.Z.

1987-01-01

To observe partial overlapping of the heat carrier cross section in piles, leading to local temperature rise or boiling of the sodium, provision is made for individual monitoring of the fuel assemblies with respect to the output temperature. Since the deviation of the mean flow rate through the pile and the output temperature is slight with this anomaly, the temperature fluctuations may provide a more informative index. The change in noise characteristics with partial overlapping of the cross sections occurs because of strong distortion of the temperature profile in the overlap region. The turbulent flow in the upper part of the pile transforms this nonuniformity into temperature pulsations which may be recorded by a sensor at the pile output. In this paper the characteristics of temperature noise are studied for various pile conditions and sensor locations by statistical modeling

Activity report of Reactor Physics Division - 1993

International Nuclear Information System (INIS)

Indira, R.

1994-01-01

The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs

Activity report of Reactor Physics Division-1995

International Nuclear Information System (INIS)

Gopalakrishnan, V.

1996-01-01

The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1995 are reported. The activity are arranged under the headings: Nuclear Data Processing and Validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. refs., figs., tabs

Activity report of Reactor Physics Division - 1993

Energy Technology Data Exchange (ETDEWEB)

Indira, R [ed.; Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

1994-12-31

The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs.

Thermal shield support degradation in pressurized water reactors

International Nuclear Information System (INIS)

Sweeney, F.J.; Fry, D.N.

1986-01-01

Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

Critical fluctuations of the number of neutrons in a reactor

International Nuclear Information System (INIS)

Ryazanov, V.V.; Lakoza, E.L.; Sysoev, V.M.

1995-01-01

The nuclear chain reaction is the most important physical process in a reactor. The theory of nuclear chain reaction fluctuations (neutron noise), developed in and other studies, has given results that are important for reactor physics and reactor practice (correlation analysis of neutron noise for measurement of the physical characteristics and reactor monitoring, stability of the critical state, etc.). Here we propose to study these problems by applying the methods of continuous phase transitions and synergetics and using the analogy with chemical chain reactions and the general laws of critical phenomena. The optimal reactor operating conditions are critical. To predict how a critical reactor will behave it is necessary to reveal those features of the neutron laws that are universal in some way, i.e., do not depend on the details of the individual acts of neutron motion and transformation that occur in reactors of different types. The similarity between chemical and nuclear chain reactions was noted long ago. Consequently, a universal theory of continuous phase transition was developed for systems of diverse physical nature

Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

Energy Technology Data Exchange (ETDEWEB)

Schuh, N J.H.

1966-12-15

This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed.

Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

International Nuclear Information System (INIS)

Schuh, N.J.H.

1966-12-01

This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed

Specialists' meeting on sodium boiling noise detection. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-01-01

The purpose of the meeting was to review and discuss methods available for detection of the initial stage of accidents in fast reactors, with most attention on reliable detection by acoustic techniques, which could provide a valuable addition to the safety protection. Results obtained from reactor experiments were also discussed and recommendations made for future developments. The meeting was divided into five technical sessions as follows: Signals from sodium boiling; Transmission of acoustic waves and background noise; Detection techniques; Reactor experiments; and Future requirements.

Specialists' meeting on sodium boiling noise detection. Summary report

International Nuclear Information System (INIS)

1982-01-01

The purpose of the meeting was to review and discuss methods available for detection of the initial stage of accidents in fast reactors, with most attention on reliable detection by acoustic techniques, which could provide a valuable addition to the safety protection. Results obtained from reactor experiments were also discussed and recommendations made for future developments. The meeting was divided into five technical sessions as follows: Signals from sodium boiling; Transmission of acoustic waves and background noise; Detection techniques; Reactor experiments; and Future requirements

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods, Stage 16

Energy Technology Data Exchange (ETDEWEB)

Pazsit, Imre; Dykin, Victor; Jonsson, Anders; Demaziere, Christophe (Chalmers Univ. of Technology, Dept. of Nuclear Engineering, Goeteborg (Sweden))

2010-12-15

This report gives an account of the work performed by the Dept. of Nuclear Engineering, Chalmers Univ. of Technology, in the frame of a research contract with the Swedish Radiation Safety Authority (SSM), contract No. SSM 2009/2093. The present report is based on work performed by Imre Pazsit, Victor Dykin, Anders Jonsson and Christophe Demaziere, with Imre Pazsit being the project leader. This report describes the results obtained during Stage 16 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The long-term goals are elaborated in more detail in e.g. the Final Reports of Stage 1 and 2 (SKI Report 95:14 and 96:50, Pazsit et al. 1995, 1996). Results up to Stage 15 were reported in (Pazsit et al. 1995, 1996, 1997, 1998, 1999, 2000, 2001, 2003a, 2003b; Demaziere et al, 2004; Sunde et al, 2006; Pazsit et al. 2008, 2009). A brief proposal for the continuation of this program in Stage 17 is also given at the end of the report. The program executed in Stage 16 consists of four parts as follows: - An overview of the present status of experience with BWR stability; - An investigation of the significance of the properties of the noise source for BWR instability; - Study of the dynamics of molten salt systems: construction of the adjoint and calculating the space dependent noise induced by propagating perturbations in the fuel; - A specific study of some novel methods of analysis of non-linear and non-stationary processes

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods, Stage 16

International Nuclear Information System (INIS)

Pazsit, Imre; Dykin, Victor; Jonsson, Anders; Demaziere, Christophe

2010-12-01

This report gives an account of the work performed by the Dept. of Nuclear Engineering, Chalmers Univ. of Technology, in the frame of a research contract with the Swedish Radiation Safety Authority (SSM), contract No. SSM 2009/2093. The present report is based on work performed by Imre Pazsit, Victor Dykin, Anders Jonsson and Christophe Demaziere, with Imre Pazsit being the project leader. This report describes the results obtained during Stage 16 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The long-term goals are elaborated in more detail in e.g. the Final Reports of Stage 1 and 2 (SKI Report 95:14 and 96:50, Pazsit et al. 1995, 1996). Results up to Stage 15 were reported in (Pazsit et al. 1995, 1996, 1997, 1998, 1999, 2000, 2001, 2003a, 2003b; Demaziere et al, 2004; Sunde et al, 2006; Pazsit et al. 2008, 2009). A brief proposal for the continuation of this program in Stage 17 is also given at the end of the report. The program executed in Stage 16 consists of four parts as follows: - An overview of the present status of experience with BWR stability; - An investigation of the significance of the properties of the noise source for BWR instability; - Study of the dynamics of molten salt systems: construction of the adjoint and calculating the space dependent noise induced by propagating perturbations in the fuel; - A specific study of some novel methods of analysis of non-linear and non-stationary processes

Activity report of Reactor Physics Division : 1990

International Nuclear Information System (INIS)

Mohanakrishnan, P.

1991-01-01

The major Research and Development and Project activities carried out during the year 1990 in Reactor Physics Division are presented in the form of summaries in this report. The various activities are organised under the following areas : (1) Nuclear Data Evaluation, Processing and Validation, (2) Core Physics and Analysis, (3) Reactor Kinetics and Safety Analysis, (4) Noise Analysis, and (5) Radiation Transport and Shielding. FBTR was restarted in July 1990 and the power was raised upto 500 kW. A number of low power physics experiments on reactivity coefficients, kinetics and noise, neutron flux and gamma dose in B cells, were performed, which are discussed in this report. (author). figs., tabs

Background noise of acoustic emission signals in sodium piping loop

International Nuclear Information System (INIS)

Mori, Y.; Aoki, K.; Kuribayashi, K.; Kishi, T.; Sakakibara, Y.

1985-01-01

Background noise measurement in the frequency range of acoustic emission (AE) signals was made on the sodium piping loops of a 50 MW steam generator test facility in the Power Reactor and Nuclear Fuel Development Corporation (PNC). During the dynamic characteristics test of the steam generator over a wide range of operating conditions, the background noise generated on the pipe surface was measured using wideband AE sensor externally mounted with waveguide. Data were obtained for the effect of power loads of steam generator on both amplitude and frequency spectra of background noise signals. Source and nature of background noise were established

Investigation of the resonant power oscillation in the Halden Boiling Water Reactor by autoregressive modeling

International Nuclear Information System (INIS)

Oguma, Ritsuo

1980-01-01

In the HBWR (Halden Boiling Water Reactor), there exists a resonant power oscillation with period about 0.04 Hz at power levels higher than about 9.5 MWt. While the resonant oscillation in not so large as to affect the normal reactor operation, it is significant, from the viewpoint of reactor diagnosis, to grasp its characteristics and find the cause. Noise analysis based on the autoregressive (AR) modeling technique has been made to reveal the driving source for this oscillation which led to the suggestion that it is attributed to the dynamic interference of heat exchange process between two parallel-connected steam transformers against the reactor. The present study demonstrates that the method used here is highly effective for tracing back to a noise source inducing the variation of quantities in a system, and also applicable to problems of reactor noise analysis and diagnosis. (author)

Kinetics, dynamics and neutron noise in Molten Salt Reactors

International Nuclear Information System (INIS)

Pazsit, Imre

2013-01-01

Reactor kinetic and dynamic properties of Molten Salt Reactors (MSR) are investigated in a simple model, which allows closed compact analytical solutions to be obtained. The goal is to gain insight, rather than to produce high-quality quantitative data. Through an interpretation of the different terms in the basic equations, and by means of analytical solutions, various approximations are introduced and their validity discussed. The dynamical behaviour of MSRs and their response to small stationary perturbations is described and discussed in comparison with traditional systems. (author)

An analytical model for studying noise effects in PWR type reactors

International Nuclear Information System (INIS)

Meyer, K.

1975-10-01

An analytical model based on the one-group diffusion method is described. It has been used for calculating the axial dependence of the spectral density of the ionization chamber noise supposing a site-independent stationary neutron flux distribution. Coolant inlet temperature fluctuations are considered as noise sources. (author)

Proceedings of the Conference on research reactors application in Yugoslavia

International Nuclear Information System (INIS)

1978-05-01

The Conference on research reactors operation was organised on the occasion of 20 anniversary of the RB zero power reactor start-up. The presentations showed that research reactors in Yugoslavia, RB, RA and TRIGA had an important role in development of nuclear sciences and technology in Yugoslavia. The reactors were applied in non-destructive testing of materials and fuel elements, development of reactor noise techniques, safety analyses, reactor control methods, neutron activation analysis, neutron radiography, dosimetry, isotope production, etc [sr

Identification of flow patterns by neutron noise analysis during actual coolant boiling in thin rectangular channels

International Nuclear Information System (INIS)

Kozma, R.; van Dam, H.; Hoogenboom, J.E.

1992-01-01

The primary objective of this paper is to introduce results of coolant boiling experiments in a simulated materials test reactor-type fuel assembly with plate fuel in an actual reactor environment. The experiments have been performed in the Hoger Onderwijs Reactor (HOR) research reactor at the Interfaculty Reactor Institute, Delft, The Netherlands. In the analysis, noise signals of self-powered neutron detectors located in the neighborhood of the boiling region and thermocouple in the channel wall and in the coolant are used. Flow patterns in the boiling coolant have been identified by means of analysis of probability density functions and power spectral densities of neutron noise. It is shown that boiling has an oscillating character due to partial channel blockage caused by steam slugs generated periodically between the plates. The observed phenomenon can serve as a basis for a boiling detection method in reactors with plate-type fuels

'Quantization' of stochastic variables: description and effects on the input noise sources in a BWR

International Nuclear Information System (INIS)

Matthey, M.

1979-01-01

A set of macrostochastic and discrete variables, with Markovian properties, is used to characterize the state of a BWR, whose input noise sources are of interest. The ratio between the auto-power spectral density (APSD) of the neutron noise fluctuations and the square modulus of the transfer function (SMTF) defines 'the total input noise source' (TINS), the components of which are the different noise source corresponding to the relevant variables. A white contribution to TINS arises from the birth and death processes of neutrons in the reactor and corresponds to a 'shot noise' (SN). Non-white contributions arise from fluctuations of the neutron cross-sections caused by fuel temperature and steam content variations. These terms called 'Flicker noises' (FN) are characterized by cut-off frequencies related to time constants of reactivity feedback effects. The respective magnitudes of the shot and flicker noises depend not only on the frequency, the feedback reactivity coefficients or the power of the reactor, but also on the 'quantization' of the continuous variables introduced such as fuel temperature and steam content. The effects of this last 'quantization' on the shapes of the noise sources and their sum are presented in this paper. (author)

Investigation of slightly forced buoyant flow in a training reactor

International Nuclear Information System (INIS)

Legradi, G.; Aszodi, A.; Por, G.

2001-01-01

A measurement based on the temperature noise analysis method was carried out in the Training Reactor of the Budapest University of Technology and Economics. The main goals were the estimation of the flow velocity immediately above the reactor core and investigation of the thermal-hydraulical conditions of the reactor, mainly in the core. Subsequently 2D and 3D computations were carried out with the aid of the code CFX- 4.3. The main objective of the 2D calculation was to clarify the thermal-hydraulical conditions of the whole reactor tank with a reasonable computing demand. It was also necessary to accomplish 3D numerical investigations of the reactor core and the space above since three dimensional effects of the flow could only be studied in this way. In addition, obtaining certain boundary conditions of the 3D computations was another significant aim of the 2D investigations. It is important that the results of the noise analysis and the operational measuring system of the reactor gave us a basis for verifying our computations.(author)

Ultra-Low Noise Germanium Neutrino Detection system (ULGeN).

Energy Technology Data Exchange (ETDEWEB)

Cabrera-Palmer, Belkis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Barton, Paul [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2017-07-01

Monitoring nuclear power plant operation by measuring the antineutrino flux has become an active research field for safeguards and non-proliferation. We describe various efforts to demonstrate the feasibility of reactor monitoring based on the detection of the Coherent Neutrino Nucleus Scattering (CNNS) process with High Purity Germanium (HPGe) technology. CNNS detection for reactor antineutrino energies requires lowering the electronic noise in low-capacitance kg-scale HPGe detectors below 100 eV as well as stringent reduction in other particle backgrounds. Existing state- of-the-art detectors are limited to an electronic noise of 95 eV-FWHM. In this work, we employed an ultra-low capacitance point-contact detector with a commercial integrated circuit preamplifier- on-a-chip in an ultra-low vibration mechanically cooled cryostat to achieve an electronic noise of 39 eV-FWHM at 43 K. We also present the results of a background measurement campaign at the Spallation Neutron Source to select the area with sufficient low background to allow a successful first-time measurement of the CNNS process.

Ultra-Low Noise Germanium Neutrino Detection system (ULGeN)

International Nuclear Information System (INIS)

Cabrera-Palmer, Belkis; Barton, Paul

2017-01-01

Monitoring nuclear power plant operation by measuring the antineutrino flux has become an active research field for safeguards and non-proliferation. We describe various efforts to demonstrate the feasibility of reactor monitoring based on the detection of the Coherent Neutrino Nucleus Scattering (CNNS) process with High Purity Germanium (HPGe) technology. CNNS detection for reactor antineutrino energies requires lowering the electronic noise in low-capacitance kg-scale HPGe detectors below 100 eV as well as stringent reduction in other particle backgrounds. Existing state- of-the-art detectors are limited to an electronic noise of 95 eV-FWHM. In this work, we employed an ultra-low capacitance point-contact detector with a commercial integrated circuit preamplifier- on-a-chip in an ultra-low vibration mechanically cooled cryostat to achieve an electronic noise of 39 eV-FWHM at 43 K. We also present the results of a background measurement campaign at the Spallation Neutron Source to select the area with sufficient low background to allow a successful first-time measurement of the CNNS process.

Fast flux test facility noise data management

International Nuclear Information System (INIS)

Thie, J.A.

1988-01-01

An extensive collection of spectra from an automated data collection system at the Fast Flux Facility has features from neutron data extracted and managed by database software. Inquiry techniques, including screening, applied to database results show the influences of control rods on wideband noise and, more generally, abilities to detect diverse types of off-normal noise. Uncovering a temporary 0.1-Hz resonance shift gave additional diagnostic information on a 13-Hz mechanical motion characterized by the interference of two resonances. The latter phenomenon is discussed generically for possible application to other reactor types. (author)

Monitoring of core barrel vibrations in WWER type reactor using out-of-reactor ionization chambers

International Nuclear Information System (INIS)

Dach, K.

1982-01-01

Vibration of the core barrel is least desirable for safe operation of the PWR reactor. These mechanical vibrations are in correlation with the fluctuations of neutron flux density whose time and frequency analysis serves failure diagnosis. The mathematical model is described of the transfer of mechanical vibrations of the core barrel to neutron noise. Other steps are indicated indispensable for the application of the method of neutron noise analysis for in-service diagnostics of nuclear power plants. (Z.M.)

Development of An Embedded FPGA-Based Data Acquisition System Dedicated to Zero Power Reactor Noise Experiments

Directory of Open Access Journals (Sweden)

Arkani Mohammad

2014-08-01

Full Text Available An embedded time interval data acquisition system (DAS is developed for zero power reactor (ZPR noise experiments. The system is capable of measuring the correlation or probability distribution of a random process. The design is totally implemented on a single Field Programmable Gate Array (FPGA. The architecture is tested on different FPGA platforms with different speed grades and hardware resources. Generic experimental values for time resolution and inter-event dead time of the system are 2.22 ns and 6.67 ns respectively. The DAS can record around 48-bit x 790 kS/s utilizing its built-in fast memory. The system can measure very long time intervals due to its 48-bit timing structure design. As the architecture can work on a typical FPGA, this is a low cost experimental tool and needs little time to be established. In addition, revisions are easily possible through its reprogramming capability. The performance of the system is checked and verified experimentally.

Study on Noise Prediction Model and Control Schemes for Substation

Science.gov (United States)

Gao, Yang; Liu, Songtao

2014-01-01

With the government's emphasis on environmental issues of power transmission and transformation project, noise pollution has become a prominent problem now. The noise from the working transformer, reactor, and other electrical equipment in the substation will bring negative effect to the ambient environment. This paper focuses on using acoustic software for the simulation and calculation method to control substation noise. According to the characteristics of the substation noise and the techniques of noise reduction, a substation's acoustic field model was established with the SoundPLAN software to predict the scope of substation noise. On this basis, 4 reasonable noise control schemes were advanced to provide some helpful references for noise control during the new substation's design and construction process. And the feasibility and application effect of these control schemes can be verified by using the method of simulation modeling. The simulation results show that the substation always has the problem of excessive noise at boundary under the conventional measures. The excess noise can be efficiently reduced by taking the corresponding noise reduction methods. PMID:24672356

Noise and its application to neutron flux measurements

International Nuclear Information System (INIS)

Sabate Puigmal, Pedro.

1984-08-01

Fission Counter's (FC) fundamental principles were studied, operating this neutron detector as pulses generator (AC modes) and fluctuant current (DC modes). Power spectral series were obtained in DC modes, corresponding to: alpha activity of the FC neutron converter, gamma exposition in Co 60 radiation field, and only neutronic field. These noise spectra were correlated with those obtained from the FC in RA-3 critical facility, at different reactor power levels. These experiments allow to verify that, in DC mode, the power noise is very weakly dependent of the reactor gamma field, over a wide range of reactor working power, and that this range is strongly dependent of the detector's position with respect to the core's position. The frequency band of measurement is not critical. The results suggest that it is possible to develop a simple and compact measurement chain for nuclear reactors control. This would be obtained with an adequate combination of the FC operation ranges in AC and DC modes. Approximately ten decades in working power would be thus controlled with this unique type of detector (Campbellian method). A locally devised commercial detector (CFPT9) was used in these tests, and several of the most useful positions of the FC were determined. Frequency band from 150 Hz to 150 KHz was investigated. (M.E.L.) [es

Activity report of Reactor Physics Section - 1985

International Nuclear Information System (INIS)

John, T.M.

1986-01-01

This Activity Report contains brief summaries of different studies made in Reactor Physics Section during the year 1985. These are presented under the headings Nuclear Data Processing and Validation, Reactor Design and Analysis, Safety and Noise Analysis, Radiation Transport and Shielding, Reactor Physics Experiments and Statistical Physics. The work on nuclear data during this period comprises primarily of validation of data of 232 Th and 233 U as a part of participation in the Co-ordinated Research Programme (CRP) under IAEA research contract. The most significant event during 1985 at this centre has been the first criticality of FBTR (Fast Breeder Test Reactor), which was achieved on the 18th of October. Reactor Physics Section has played a key role in this event by carrying out the first approach to criticality with fuel loading in a safe manner and conducting some low power reactor physics experiments which are discussed. The studies made in the field reactor safety and shielding are also connected mainly with the FBTR problems in addition to some work on the PFBR (Prototype Fast Breeder Reactor) detailed design of which has been just started. Studies pertaining to the other two Co-ordinated Research Programmes (CRP) under IAEA contract, namely (1) on the comparative assessment of processing techniques for the analysis of sodium boiling noise detection and, (2) on the contribution of advanced reactors to energy supply have been continued during this year. At the end of this report, a list of publications made by the members of the section and also the sectional seminars held during this period is included. (author)

Research and development program in reactor diagnostics and monitoring with neutron noise methods. Stage 7. Final report

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Demaziere, C.; Arzhanov, V. [Chalmers Univ. of Technology, Goeteborg (Sweden). Department of Reactor Physics; Garis, N.S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

2001-08-01

This report constitutes stage 7 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. A proposal for the continuation of this program in stage 8 is also given at the end of the report. In stage 6, the basic principles of a 3-D fully coupled neutronic/thermal-hydraulic simulator in the frequency domain were presented. The neutronic model relied on the two-group diffusion approximation, whereas the thermal-hydraulic algorithms relied on the so called 'lumped' model. The key element of this simulator was that only the static data were required which could be obtained from the Studsvik Scandpower CASMO-4/TABLES-3/ SIMULATE-3 code package. The simulator was developed with this underlying idea, which means that the calculation of the static fluxes and the eigenvalue were avoided. Depending on what kind of spatial discretization scheme which is used in the noise simulator to calculate the 'leakage' noise, it is not granted that the system remains critical by using the group constants supplied by SIMULATE. Nevertheless, when the system is critical, the balance equations should be fulfilled in all nodes with respect to the discretization scheme used. In concrete terms, the calculation of the static fluxes and eigenvalue can be avoided if the system is brought back to criticality by modifying the cross-sections so that the balance equations are always fulfilled with the chosen spatial discretization scheme. This approach was used in this study with the finite difference scheme. As pointed out in stage 6, the finite difference scheme is relatively inefficient compared to finite elements or nodal methods, but on the other hand it is rather easy to implement. These two more sophisticated schemes are planned to be investigated at a later stage, but for the time being the simulator relying on the finite difference scheme was improved as much as possible so that a 2-D entirely

Noise diagnosis - a method for early detection of failures in a nuclear plant

International Nuclear Information System (INIS)

Brinckmann, H.F.

1981-01-01

Noise diagnosis constitutes one method for early detection of plant failures. The method is based on the fact that nearly all undesired processes in a nuclear power plant make a measurable contribution to the noise portion of signals. Well-known examples of undesired processes in pressurized water reactors include core-barrel movement, the vibration of control elements, the appearance of loose parts in the coolant flow, and the process of coolant boiling. Each of these processes has been implicated in past nuclear plant failures. In the German Democratic Republic (GDR) P. Liewers and his colleagues have introduced noise analysis systems into the primary circuit of WWER-440 pressurized water reactors (PWR). The most progressive version (RAS-II) has become a prototype for research and routine investigations. This system is described. (author)

Real-time stability monitoring method for boiling water reactor nuclear power plants

International Nuclear Information System (INIS)

Fukunishi, K.; Suzuki, S.

1987-01-01

A method for real-time stability monitoring is developed for supervising the steady-state operation of a boiling water reactor core. The decay ratio of the reactor power fluctuation is determined by measuring only the output neutron noise. The concept of an inverse system is introduced to identify the dynamic characteristics of the reactor core. The adoption of an adaptive digital filter is useful in real-time identification. A feasibility test that used measured output noise as an indication of reactor power suggests that this method is useful in a real-time stability monitoring system. Using this method, the tedious and difficult work for modeling reactor core dynamics can be reduced. The method employs a simple algorithm that eliminates the need for stochastic computation, thus making the method suitable for real-time computation with a simple microprocessor. In addition, there is no need to disturb the reactor core during operation. Real-time stability monitoring using the proposed algorithm may allow operation under less stable margins

Relationship of core exit-temperature noise to thermal-hydraulic conditions in PWRs

International Nuclear Information System (INIS)

Sweeney, F.J.; Upadhyaya, B.R.

1983-01-01

Core exit thermocouple temperature noise and neutron detector noise measurements were performed at the Loss of Fluid Test Facility (LOFT) reactor and a Westinghouse, 1148 MW(e) PWR to relate temperature noise to core thermal-hydraulic conditions. The noise analysis results show that the RMS of the temperature noise increases linearly with increasing core δT at LOFT and the commercial PWR. Out-of-core test loop temperature noise has shown similar behavior. The phase angle between core exit temperature noise and in-core or ex-core neutron noise is directly related to the core coolant flow velocity. However, if the thermocouple response time is slow, compared to the coolant transit time between the sensors, velocities inferred from the phase angle are lower than measured coolant flow velocities

NOASYS, a system for on-line noise analysis

International Nuclear Information System (INIS)

Massier, H.

1978-07-01

This report describes NOASYS, a versatile NOise Analysis SYStem for digital on-line signal processing. The system based on a minicomputer, was originally developed for the analysis of noise signals from nuclear reactors. NOASYS computes various statistical functions e.g. cross- und auto-correlation functions and power spectral densities resp., which may be used for reactor diagnosis and malfunction detection. The system processes up to 16 analog signals with a maximum sampling frequency of 100 kcps (1 channel). The processing of the sampled data is done by a number of software tasks, which may be called from a teletype or linked together for specific measuring programs. The standard configuration contains the often used processing routines e.g. Fast Fourier Transform, complex multiplication, summation etc. In addition NOASYS may be extended by users own tasks (in Assembler or Fortran) to fit for specific applications. (orig.) 891 HP [de

The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

International Nuclear Information System (INIS)

Carey, W.M.; Gavin, A.P.; Bobis, J.P.; Sheen, S.H.; Anderson, T.T.; Doolittle, R.D.; Albrecht, R.W.

1977-01-01

Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapour bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapour bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low ( 0 C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary. (author)

Temperature fluctuations: an assessment of their use in the detection of fast reactor coolant blockages

International Nuclear Information System (INIS)

Greef, C.P.

1979-01-01

The temperature noise technique for the detection of local blockages in fast reactor subassemblies is discussed. The main factors involved in an assessment of the technique are outlined and the experimental and theoretical work that has been carried out at BNL on the various aspects of the problem is described. It is concluded that blockings appreciably smaller than those predicted to produce boiling should be detectable against a background noise level due to subassembly power tilts, on a time scale giving protection against rapidly developing incidents. Further work required to increase confidence in the application of the technique to the reactor is outlined, including measurements in fully representative geometries, data from sodium rigs and further information on reactor background noise levels. (Auth.)

Count-to-count time interval distribution analysis in a fast reactor

International Nuclear Information System (INIS)

Perez-Navarro Gomez, A.

1973-01-01

The most important kinetic parameters have been measured at the zero power fast reactor CORAL-I by means of the reactor noise analysis in the time domain, using measurements of the count-to-count time intervals. (Author) 69 refs

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 11 and 12. Final report

Energy Technology Data Exchange (ETDEWEB)

Sunde, C.; Pazsit, I.; Demaziere, C.; Dahl, O.; Mileshina, L. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Engineering

2006-06-15

This report describes the results obtained during Stage 11 and 12 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The long-term goals are elaborated in more detail in e.g. the Final Reports of stage 1 and 2 (SKI Report 95:14 and 96:50). Higher order eigenfunctions and eigenvalues of the diffusion equation, describing a static core, have lost their significance when doing calculations in realistic systems, since there are usually made by nodal methods or other direct numerical techniques. However, there are situations when knowledge of the higher order modes is still useful. Such case is the separation of the global and regional flux oscillations in the case of BWR instability. Another case is the investigation of the stability properties of large inhomogeneous cores, which is usually quantified with the so-called eigenvalue separation, ES=1/k{sub 1} - 1/k{sub 0} where k{sub 0} and k{sub 1}, are the fundamental and first higher order eigenvalues, respectively. Numerical codes used for ICMF calculations usually do not have the option of calculating the higher order eigenvalues and eigenfunctions. In idealised systems, these can be calculated with semi-analytical methods. In order to get insight into the characteristics of the higher order eigenmodes of the neutron flux and the adjoint, these were calculated in a reflected homogeneous system with two-group theory. The adjoint eigenmodes are necessary if an orthogonality property needs to be used, such as the separation of the modes from a flux shape which is a sum of several eigenmodes. At the same time, the so-called numerical noise simulator, developed at the Dept., was extended such that it became suitable for the calculation of the higher order eigenmodes, both direct and adjoint ones. This simulator can treat real inhomogeneous cores, with an input deck compatible with that of SIMULATE. The simulator was used to calculate the

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 11 and 12. Final report

International Nuclear Information System (INIS)

Sunde, C.; Pazsit, I.; Demaziere, C.; Dahl, O.; Mileshina, L.

2006-06-01

This report describes the results obtained during Stage 11 and 12 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The long-term goals are elaborated in more detail in e.g. the Final Reports of stage 1 and 2 (SKI Report 95:14 and 96:50). Higher order eigenfunctions and eigenvalues of the diffusion equation, describing a static core, have lost their significance when doing calculations in realistic systems, since there are usually made by nodal methods or other direct numerical techniques. However, there are situations when knowledge of the higher order modes is still useful. Such case is the separation of the global and regional flux oscillations in the case of BWR instability. Another case is the investigation of the stability properties of large inhomogeneous cores, which is usually quantified with the so-called eigenvalue separation, ES=1/k 1 - 1/k 0 where k 0 and k 1 , are the fundamental and first higher order eigenvalues, respectively. Numerical codes used for ICMF calculations usually do not have the option of calculating the higher order eigenvalues and eigenfunctions. In idealised systems, these can be calculated with semi-analytical methods. In order to get insight into the characteristics of the higher order eigenmodes of the neutron flux and the adjoint, these were calculated in a reflected homogeneous system with two-group theory. The adjoint eigenmodes are necessary if an orthogonality property needs to be used, such as the separation of the modes from a flux shape which is a sum of several eigenmodes. At the same time, the so-called numerical noise simulator, developed at the Dept., was extended such that it became suitable for the calculation of the higher order eigenmodes, both direct and adjoint ones. This simulator can treat real inhomogeneous cores, with an input deck compatible with that of SIMULATE. The simulator was used to calculate the eigenfunctions and

Detection of boiling by Piety's on-line PSD-pattern recognition algorithm applied to neutron noise signals in the SAPHIR reactor

International Nuclear Information System (INIS)

Spiekerman, G.

1988-09-01

A partial blockage of the cooling channels of a fuel element in a swimming pool reactor could lead to vapour generation and to burn-out. To detect such anomalies, a pattern recognition algorithm based on power spectra density (PSD) proposed by Piety was further developed and implemented on a PDP 11/23 for on-line applications. This algorithm identifies anomalies by measuring the PSD on the process signal and comparing them with a standard baseline previously formed. Up to 8 decision discriminants help to recognize spectral changes due to anomalies. In our application, to detect boiling as quickly as possible with sufficient sensitivity, Piety's algorithm was modified using overlapped Fast-Fourier-Transform-Processing and the averaging of the PSDs over a large sample of preceding instantaneous PSDs. This processing allows high sensitivity in detecting weak disturbances without reducing response time. The algorithm was tested with simulation-of-boiling experiments where nitrogen in a cooling channel of a mock-up of a fuel element was injected. Void fractions higher than 30 % in the channel can be detected. In the case of boiling, it is believed that this limit is lower because collapsing bubbles could give rise to stronger fluctuations. The algorithm was also tested with a boiling experiment where the reactor coolant flow was actually reduced. The results showed that the discriminant D5 of Piety's algorithm based on neutron noise obtained from the existing neutron chambers of the reactor control system could sensitively recognize boiling. The detection time amounts to 7-30 s depending on the strength of the disturbances. Other events, which arise during a normal reactor run like scrams, removal of isotope elements without scramming or control rod movements and which could lead to false alarms, can be distinguished from boiling. 49 refs., 104 figs., 5 tabs

Digital signal processing for the Johnson noise thermometry: a time series analysis of the Johnson noise

International Nuclear Information System (INIS)

Moon, Byung Soo; Hwang, In Koo; Chung, Chong Eun; Kwon, Kee Choon; David, E. H.; Kisner, R.A.

2004-06-01

In this report, we first proved that a random signal obtained by taking the sum of a set of signal frequency signals generates a continuous Markov process. We used this random signal to simulate the Johnson noise and verified that the Johnson noise thermometry can be used to improve the measurements of the reactor coolant temperature within an accuracy of below 0.14%. Secondly, by using this random signal we determined the optimal sampling rate when the frequency band of the Johnson noise signal is given. Also the results of our examination on how good the linearity of the Johnson noise is and how large the relative error of the temperature could become when the temperature increases are described. Thirdly, the results of our analysis on a set of the Johnson noise signal blocks taken from a simple electric circuit are described. We showed that the properties of the continuous Markov process are satisfied even when some channel noises are present. Finally, we describe the algorithm we devised to handle the problem of the time lag in the long-term average or the moving average in a transient state. The algorithm is based on the Haar wavelet and is to estimate the transient temperature that has much smaller time delay. We have shown that the algorithm can track the transient temperature successfully

Activity report of Reactor Physics Division - 1988

International Nuclear Information System (INIS)

Keshavamurthy, R.S.

1989-01-01

This report highlights the progress of activities carried out during the year 1988 in Reactor Physics Division in the form of brief summaries. The topics are organised under the following subject categories:(1) nuclear data evaluation , processing and validation, (2) core physics and analysis, (3) reactor kinetics and safety analysis, (4) noise analysis and (5) radiation transport and shielding. List of publications by the members of the Division and the Reactor Physics Seminars held during the year 1988, is included at the end of report. (author). refs., figs., tabs

'Kazmer' a complex noise diagnostic system for 1000 MWe PWR WWER type nuclear power units

International Nuclear Information System (INIS)

Por, G.

1992-06-01

Noise diagnostic systems have previously been developed and installed for the WWER-440 type reactors at the Paks Nuclear Power Plant, Hungary. Based on the experiences, the system has been extended and modified for use in 1000 MWe, WWER-1000 type units. KAZMER consists of three subsystem, the KARD reactor noise diagnostic system, ARGUS vibration monitoring system for rotation machinery, and ALMOS acoustic monitoring system. The installation of the KAZMER system at the Kalinin Nuclear Power Station, Russia, and the first operational experiences are outlined. (R.P.) 15 refs.; 9 figs

Determination of fuel assembly vibrational modes through analysis of incore detector noise

International Nuclear Information System (INIS)

Johnson, R.S.

1986-01-01

In order to better characterize fuel assembly vibration at Duke Power Company's Oconee Nuclear Station, incore noise data were acquired an analyzed from prompt responding detectors incorporated in the Oconee 2, Cycle 7 core. Duke Power Company began actively pursuing an inhouse Neutron Noise Analysis program for routine surveillance of reactor internals vibration in 1979. Noise data has since been acquired and analyzed for twelve cycles of operation for the three Oconee units. Duke Power's Oconee Unit 2 is a Babcock and Wilcoxs pressurized water reactor with a rate thermal power of 2568MW. For Oconee 2, Cycle 7 operation, two test assemblies, each employing a string of seven axially-spaced, prompt responding hafnium detectors, were included in the final core design. Incore detector noise data were obtained during Cycle 7 at approximately 281 and 430 effective full power days (EFPD). In addition to the incore test detector signals, noise signals from the upper and lower chambers of the four excore power range detectors were recorded to aid in the analysis. The comparison of RMS signal levels for each incore detector and the phase relationships between detector locations within two test assemblies identified the first four fuel assembly bending modes associated with fixed end conditions

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)

Development of a real-time stability measurement system for boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; King, W.T.

1987-01-01

This paper describes the development of a portable, real time system for boiling water reactor (BWR) stability measurements. The system provides a means for the operator to monitor the reactor stability using existing plant instrumentation and commercially available hardware. The noise component (i.e., perturbations around steady state) of the neutron signal in BWRs has been shown to contain information about reactor stability, and several algorithms have been developed to extract that information. For the present work, the authors have used an algorithm that has been implemented on a portable personal computer. This algorithm uses the autocorrelation function of naturally occurring neutron noise (measured without special plant perturbations) and an autoregressive modeling technique to produce the asymptotic DR. For this real-time implementation, neutron noise data is preconditioned (i.e., filtered and amplified) and sampled at a 5-Hz sampling rate using a commercial data-acquisition system. Approximately every 1.5 min, the current (snapshot) autocorrelation function is computed directly from the data, and the average autocorrelation is updated. The current and average DR estimates are evaluated with the same periodicity and are displayed on the screen along with the autocorrelations and average power spectrum of the neutron noise

Stability monitoring of a natural-circulation-cooled boiling water reactor

International Nuclear Information System (INIS)

Hagen, T.H.J.J. van der.

1989-01-01

Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

Count-to-count time interval distribution analysis in a fast reactor; Estudio de la distribucion de intervalos de tiempo entre detecciones consecutivas de neutrones en un reactor rapido

Energy Technology Data Exchange (ETDEWEB)

Perez-Navarro Gomez, A

1973-07-01

The most important kinetic parameters have been measured at the zero power fast reactor CORAL-I by means of the reactor noise analysis in the time domain, using measurements of the count-to-count time intervals. (Author) 69 refs.

Measuring temperature coefficient of TRIGA MARK I reactor by noise analysis

International Nuclear Information System (INIS)

Soares, P.A.

1975-01-01

The transfer function of TRIGA MARK I Reactor is measured at power zero (5w) and power 118Kw, in the frequency range of 0.02 to 0.5 rd/s. The method of intercorrelation between a pseudostochasticbinary signal is used. A simple dynamic model of the reactor is developed and the coefficient of temperature is estimated [pt

Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

Nuclear reactor noise investigations on boiling effects in a simulated MTR-type fuel assembly

International Nuclear Information System (INIS)

Kozma, R.

1992-01-01

The work includes validation/testing of existing neutron noise methods under well-controlled circumstances, investigation of boiling phenomena in narrow channels, and development of a novel boiling monitoring method. The work has been performed in the NIOBE facility at the HDR. Noise signals of thermocouples in the channel wall are used for velocity profile monitoring. Flow patterns in the boiling coolant are identified by means of analysis of probaof probability density functions and neutron noise spectra. Local noise effects are studied. (DG)

High temperature measurement by noise thermometry

International Nuclear Information System (INIS)

Decreton, M.C.

1982-06-01

Noise thermometry has received a lot of attention for measurements of temperatures in the high range around 1000-2000 deg. K. For these measurements, laboratory type experiments have been mostly performed. These have shown the interest of the technique when long term stability, high precision and insensibility to external conditions are concerned. This is particularly true for measurements in nuclear reactors where important drifts due to irradiation effects are experienced with other measurement techniques, as thermocouple for instance. Industrial noise thermometer experiments have not been performed extensively up to now. The subject of the present study is the development of a 1800 deg. K noise thermometer for nuclear applications. The measurement method is based on a generalized noise power approach. The rms noise voltage (Vsub(s)) and noise current (Isub(s)) are successively measured on the resistive sensor. The same quantities are also measured on a dummy short circuited probe (Vsub(d) and Isub(d)). The temperature is then deduced from these measured values by the following formula: cTsub(s) = (Vsub(s) 2 - Vsub(d) 2 )(Vsub(s)/Isub(s) - Vsub(d)/Isub(d)) - 1 , where c is a constant and Tsub(s) the absolute temperature of the sensor. This approach has the particular advantage of greatly reducing the sensibility to environmental perturbations on the leads and to the influence of amplifier noise sources. It also eliminates the necessity of resistance measurement and keeps the electronic circuits as simple as possible

Sensor response time monitoring using noise analysis

International Nuclear Information System (INIS)

Hashemian, H.M.; Thie, J.A.; Upadhyaya, B.R.; Holbert, K.E.

1988-01-01

Random noise techniques in nuclear power plants have been developed for system surveillance and for analysis of reactor core dynamics. The noise signals also contain information about sensor dynamics, and this can be extracted using frequency, amplitude and time domain analyses. Even though noise analysis has been used for sensor response time testing in some nuclear power plants, an adequate validation of this method has never been carried out. This paper presents the results of limited work recently performed to examine the validity of the noise analysis for sensor response time testing in nuclear power plants. The conclusion is that noise analysis has the potential for detecting gross changes in sensor response but it cannot be used for reliable measurement of response time until more laboratory and field experience is accumulated. The method is more advantageous for testing pressure sensors than it is for temperature sensors. This is because: 1) for temperature sensors, a method called Loop Current Step Response test is available which is quantitatively more exact than noise analysis, 2) no method currently exists for on-line testing of pressure transmitters other than the Power-Interrupt test which is applicable only to force balance pressure transmitters, and 3) pressure sensor response time is affected by sensing line degradation which is inherently taken into account by testing with noise analysis. (author)

Proposal to negotiate an amendment to an existing contract for the supply of four additional low-noise Thyristor-Controlled Reactor coils for a new Static VAR Compensator on the 18 kV electrical network on the Meyrin site

CERN Document Server

2017-01-01

Proposal to negotiate an amendment to an existing contract for the supply of four additional low-noise Thyristor-Controlled Reactor coils for a new Static VAR Compensator on the 18 kV electrical network on the Meyrin site

Contraction of information and its inverse problem in reactor system identification and stochastic diagnosis

International Nuclear Information System (INIS)

Kishida, K.

1996-01-01

Research concerning power reactor noise analysis makes rapid progress in the areas of the system identification, prediction and diagnosis. Keywords in these studies are artificial intelligence, neural network, fuzzy, and chaos. Nonlinear, nonstationary, or non-Gaussian processes as well as linear and steady processes are also studied in fluctuation analysis. However, we have not enough time to study a fundamental theory, since we are urged to obtain results or applications in power reactor fluctuations. Furthermore, we have no systematic approach to handle observed time series data in the linear process, since power reactor noise phenomena are complicated. Hence, it is important to study it from the fundamental viewpoint. It is a main aim of the present review paper to describe a unified formalism for reactor system identification and stochastic diagnosis

Method of starting internal pumps of a nuclear reactor

International Nuclear Information System (INIS)

Kumagami, Shoji.

1985-01-01

Purpose: To reduce the noise effects by decreasing the invading current into the main line upon starting an internal pump type nuclear reactor adapted to forcively recycle the reactor water by a plurality of internal pumps. Method: A plurality of internal pumps are divided into several groups and, upon starting pumps belonging to the individual unit group, the starting instances for the respective pumps are deviated to reduce the surges applied to the main line and suppress the invading current lower to reduce the earth noises. As a result, effects caused to other devices or equipments can be moderated to improve the reliability. Furthermore, by actuating the respective pumps on every group units in a starting pattern along the orthogonal line, flow rate distribution in the reactor can be balanced. Then, the instability region during low rotation of pumps, that is, instability of the flow rate near the resonance frequency can be decreased. (Kawakami, Y.)

Study of the noise propagation in PWR with coupled codes

International Nuclear Information System (INIS)

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

2011-01-01

The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

FFTF primary system transition to natural circulation from low reactor power

International Nuclear Information System (INIS)

Bouchey, G.D.; Additon, S.L.; Nutt, W.T.

1980-01-01

Plans for reactor and primary loop natural circulation testing in the Fast Flux Test Facility (FFTF) are summarized. Detailed pretest planning with an emphasis on understanding the implications of process noise and model uncertainties for model verification and test acceptance are discussed for a transition to natural circulation in the reactor core and primary heat transport loops from initial conditions of 5% of rated reactor power and 75% of full flow

In-situ calibration of nuclear-plant platinum resistance thermometers using Johnson noise methods. Final report

International Nuclear Information System (INIS)

Blalock, T.V.; Shepard, R.L.

1983-06-01

Methods for in situ calibration of reactor plant platinum resistance thermometers using Johnson noise measurements were tested in the laboratory and in two operating reactors: Diablo Canyon and Sequoyah. The Johnson noise methods provide an absolute measurement of the thermometer temperature and can be compared with the dc calibration of the thermometers to confirm their calibration without removing the thermometers from the plant coolant loops. Inaccuracies of less than 0.1% were obtained with these methods for ideal situations where the noise measuring equipment could be connected directly to the installed thermometer terminals. For measurements made at the ends of long extension cables, inaccuracies were 0.5 to 1.0%, at best. Extension cables must be optimized and well characterized electrically to achieve such accuracies. Other factors that affect the accuracy of these methods were evaluated

Signal processing techniques for sodium boiling noise detection

International Nuclear Information System (INIS)

1989-05-01

At the Specialists' Meeting on Sodium Boiling Detection organized by the International Working Group on Fast Reactors (IWGFR) of the International Atomic Energy Agency at Chester in the United Kingdom in 1981 various methods of detecting sodium boiling were reported. But, it was not possible to make a comparative assessment of these methods because the signal condition in each experiment was different from others. That is why participants of this meeting recommended that a benchmark test should be carried out in order to evaluate and compare signal processing methods for boiling detection. Organization of the Co-ordinated Research Programme (CRP) on signal processing techniques for sodium boiling noise detection was also recommended at the 16th meeting of the IWGFR. The CRP on Signal Processing Techniques for Sodium Boiling Noise Detection was set up in 1984. Eight laboratories from six countries have agreed to participate in this CRP. The overall objective of the programme was the development of reliable on-line signal processing techniques which could be used for the detection of sodium boiling in an LMFBR core. During the first stage of the programme a number of existing processing techniques used by different countries have been compared and evaluated. In the course of further work, an algorithm for implementation of this sodium boiling detection system in the nuclear reactor will be developed. It was also considered that the acoustic signal processing techniques developed for boiling detection could well make a useful contribution to other acoustic applications in the reactor. This publication consists of two parts. Part I is the final report of the co-ordinated research programme on signal processing techniques for sodium boiling noise detection. Part II contains two introductory papers and 20 papers presented at four research co-ordination meetings since 1985. A separate abstract was prepared for each of these 22 papers. Refs, figs and tabs

Neutron noise analysis of BWR using time series analysis

International Nuclear Information System (INIS)

Fukunishi, Kohyu

1976-01-01

The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW) [de

The Effect of Different Perturbations on the Stability Analysis of Light Water Reactors

International Nuclear Information System (INIS)

Dykin, Victor

2010-09-01

Neutron noise analysis techniques are studied and developed, with primary use of determining the stability of Boiling Water Reactors (BWRs). In particular, the role of a specific perturbation prevailing in Light Water Reactors, the propagating density perturbation, in the stability of BWRs and on the noise field of LWRs in general, is investigated by considering three topics. In the first topics, we investigate how the neutronic response of the reactor, usually described as a second order system driven by a white noise driving force, is affected by a non-white driving force. This latter arises from the reactivity effect of the propagating density perturbations. The investigation is performed by using spectral and correlation analysis. Propagating perturbations with different velocities are analyzed. We investigate how the accuracy of the determination of the so-called decay ratio (DR) of the system, based on the assumption of white noise driving force, deteriorates with deviations from the white noise character of the driving force. In the second topics, the space dependence of the neutron noise, induced by propagating density perturbations, represented through the perturbation of the absorption, is determined and discussed. A full analytical solution was obtained by the use of the Green's function technique. The solution was analyzed for different frequencies and different system sizes. An interesting new interference effect between the point-kinetic and space-dependent components of the induced noise was discovered and interpreted in physical terms. In the last topics, a non-linear stability analysis of a BWR is performed, using so called Reduced Order Model (ROM) techniques. A ROM is usually constructed by reducing the full set of 3D space-time dependent neutron-kinetics, thermal-hydraulics and heat transfer equations to time-dependent ones, by considering space dependence in a lumped parameter model (one or two discrete channels). The main novelty of our work

Treatment of the response of a reactor to stochastic reactivity input

International Nuclear Information System (INIS)

Bansal, N.K.

1977-08-01

One of the important applications of reactor noise theory, which relies on the methematical methods for treating stochastic processes, is to determine either the confidence limits for the allowed deviations of the measured signals during normal reactor operation, or the statistical properties of their respective expectation values. In this report, we stress mainly the general mathematical aspects for treating this problem. A global description of a reactor system, perturbed by stochastic reactivity input, leads to a stochastic differential equation with parametric excitation. A discrepancy exists in literature about obtaining the correct solution of such an equation in its general frame. We discuss this discrepancy and review the work done for solving such an equation. Some recent work indicates that linearisation of system's equations is justified in most cases of reactor operations. We develop a general scheme for calculating the various covariances and correlation functions in a stable and stationary system, which is perturbed by various noise sources and where linearisation of system's equations is justified. The formulation is easily extendable to an unstable, nonstationary system, like an uncontrolled critical reactor as demonstrated. (orig.) [de

Parameter identification in a nonlinear nuclear reactor model using quasilinearization

International Nuclear Information System (INIS)

Barreto, J.M.; Martins Neto, A.F.; Tanomaru, N.

1980-09-01

Parameter identification in a nonlinear, lumped parameter, nuclear reactor model is carried out using discrete output power measurements during the transient caused by an external reactivity change. In order to minimize the difference between the model and the reactor power responses, the parameter promt neutron generation time and a parameter in fuel temperature reactivity coefficient equation are adjusted using quasilinearization. The influences of the external reactivity disturbance, the number and frequency of measurements and the measurement noise level on the method accuracy and rate of convergence are analysed through simulation. Procedures for the design of the identification experiments are suggested. The method proved to be very effective for low level noise measurements. (Author) [pt

Theoretical investigation of the neutron noise diagnostics of two-dimensional control rod vibrations in a PWR

International Nuclear Information System (INIS)

Pazsit, I.; Analytis, G.T.

1980-01-01

In order to develop a method for monitoring control rod vibrations by neutron noise measurements, the noise induced by two-dimensional vibrations of control elements is investigated. The two-dimensional Green's function relating the small stochastic cross-section fluctuations to the neutron noise is determined for a rectangular slab reactor in the modified one-group theory, and subsequently, the neutron response to two-dimensional vibrating noise sources is investigated. Two possible diagnostical applications are considered: (a) the reconstruction of the mechanical trajectory of the vibrating element by neutron noise measurements, and (b) the possibility of locating the vibrating element in the core. (author)

Filtering Performance Comparison of Kernel and Wavelet Filters for Reactivity Signal Noise

International Nuclear Information System (INIS)

Park, Moon Ghu; Shin, Ho Cheol; Lee, Yong Kwan; You, Skin

2006-01-01

Nuclear reactor power deviation from the critical state is a parameter of specific interest defined by the reactivity measuring neutron population. Reactivity is an extremely important quantity used to define many of the reactor startup physics parameters. The time dependent reactivity is normally determined by solving the using inverse neutron kinetics equation. The reactivity computer is a device to provide an on-line solution of the inverse kinetics equation. The measurement signal of the neutron density is normally noise corrupted and the control rods movement typically gives reactivity variation with edge signals like saw teeth. Those edge regions should be precisely preserved since the measured signal is used to estimate the reactivity wroth which is a crucial parameter to assure the safety of the nuclear reactors. In this paper, three kind of edge preserving noise filters are proposed and their performance is demonstrated using stepwise signals. The tested filters are based on the unilateral, bilateral kernel and wavelet filters which are known to be effective in edge preservation. The bilateral filter shows a remarkable improvement compared with unilateral kernel and wavelet filters

Fast neutron reactor noise analysis: beginning failure detection and physical parameter estimation

International Nuclear Information System (INIS)

Le Guillou, G.

1975-01-01

The analysis of the signals fluctuations coming from a power nuclear reactor (a breeder), by correlation methods and spectral analysis has two principal applications: on line estimation of physical parameters (reactivity coefficients); beginning failures (little boiling, abnormal mechanic vibrations). These two applications give important informations to the reactor core control and permit a good diagnosis [fr

On the neutron noise diagnostics of pressurized water reactor control rod vibrations. 1. periodic vibrations

International Nuclear Information System (INIS)

Pazsit, I.; Glockler, O.

1983-01-01

Based on the theory of neutron noise arising from the vibration of a localized absorber, the possibility of rod vibration diagnostics is investigated. It is found that noise source characteristics, namely rod position and vibration trajectory and spectra, can be unfolded from measured neutron noise signals. For the localization process, the first and more difficult part of the diagnostics, a procedure is suggested whose novelty is that it is applicable in case of arbitrary vibration trajectories. Applicability of the method is investigated in numerical experiments where effects of background noise are also accounted for

Process sensors characterization based on noise analysis technique and artificial intelligence

International Nuclear Information System (INIS)

Mesquita, Roberto N. de; Perillo, Sergio R.P.; Santos, Roberto C. dos

2005-01-01

The time response of pressure and temperature sensors from the Reactor Protection System (RPS) is a requirement that must be satisfied in nuclear power plants, furthermore is an indicative of its degradation and its remaining life. The nuclear power industry and others have been eager to implement smart sensor technologies and digital instrumentation concepts to reduce manpower and effort currently spent on testing and calibration. Process parameters fluctuations during normal operation of a reactor are caused by random variations in neutron flux, heat transfer and other sources. The output sensor noise can be considered as the response of the system to an input representing the statistical nature of the underlying process which can be modeled using a time series model. Since the noise signal measurements are influenced by many factors, such as location of sensors, extraneous noise interference, and randomness in temperature and pressure fluctuation - the quantitative estimate of the time response using autoregressive noise modeling is subject to error. This technique has been used as means of sensor monitoring. In this work a set of pressure sensors installed in one experimental loop adapted from a flow calibration setup is used to test and analyze signals in a new approach using artificial intelligence techniques. A set of measurements of dynamic signals in different experimental conditions is used to distinguish and identify underlying process sources. A methodology that uses Blind Separation of Sources with a neural networks scheme is being developed to improve time response estimate reliability in noise analysis. (author)

Process sensors characterization based on noise analysis technique and artificial intelligence

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Roberto N. de; Perillo, Sergio R.P.; Santos, Roberto C. dos [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: rnavarro@ipen.br; sperillo@ipen.br; rcsantos@ipen.br

2005-07-01

The time response of pressure and temperature sensors from the Reactor Protection System (RPS) is a requirement that must be satisfied in nuclear power plants, furthermore is an indicative of its degradation and its remaining life. The nuclear power industry and others have been eager to implement smart sensor technologies and digital instrumentation concepts to reduce manpower and effort currently spent on testing and calibration. Process parameters fluctuations during normal operation of a reactor are caused by random variations in neutron flux, heat transfer and other sources. The output sensor noise can be considered as the response of the system to an input representing the statistical nature of the underlying process which can be modeled using a time series model. Since the noise signal measurements are influenced by many factors, such as location of sensors, extraneous noise interference, and randomness in temperature and pressure fluctuation - the quantitative estimate of the time response using autoregressive noise modeling is subject to error. This technique has been used as means of sensor monitoring. In this work a set of pressure sensors installed in one experimental loop adapted from a flow calibration setup is used to test and analyze signals in a new approach using artificial intelligence techniques. A set of measurements of dynamic signals in different experimental conditions is used to distinguish and identify underlying process sources. A methodology that uses Blind Separation of Sources with a neural networks scheme is being developed to improve time response estimate reliability in noise analysis. (author)

Measurements for kinetic parameters estimation in the RA-0 research reactor

International Nuclear Information System (INIS)

Gomez, A; Bellino, P A

2012-01-01

In the present work, measurements based on the neutron noise technique and the inverse kinetic method were performed to estimate the different kinetic parameters of the reactor in its critical state. By means of the neutron noise technique, we obtained the current calibration factor of the ionization chamber M6 belonging to the power range channels of the reactor instrumentation. The maximum current allowed compatible with the maximum power authorized by the operation license was also obtained. Using the neutron noise technique, the reduced mean reproduction time (Λ*) was estimated. This parameter plays a fundamental role in the deterministic analysis of criticality accidents. Comparison with previous values justified performing new measurements to study systematic trends in the value of Λ*. Using the inverse kinetics method, the reactivity worth of the control rods was estimated, confirming the existence of spatial effects and trends previously observed (author)

Pile noise experiment in MINERVE reactor to estimate kinetic parameters using various data processing methods

Energy Technology Data Exchange (ETDEWEB)

Geslot, Benoit; Gruel, Adrien; Pepino, Alexandra; Di Salvo, Jacques; Izarra, Gregoire de; Jammes, Christian; Destouches, Christophe; Blaise, Patrick [CEA, DEN, DER/SPEx, Cadarache, F-13108 St Paul Lez Durance (France)

2015-07-01

MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from two high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)

ARMA modelling of neutron stochastic processes with large measurement noise

International Nuclear Information System (INIS)

Zavaljevski, N.; Kostic, Lj.; Pesic, M.

1994-01-01

An autoregressive moving average (ARMA) model of the neutron fluctuations with large measurement noise is derived from langevin stochastic equations and validated using time series data obtained during prompt neutron decay constant measurements at the zero power reactor RB in Vinca. Model parameters are estimated using the maximum likelihood (ML) off-line algorithm and an adaptive pole estimation algorithm based on the recursive prediction error method (RPE). The results show that subcriticality can be determined from real data with high measurement noise using much shorter statistical sample than in standard methods. (author)

Coherence method of identifying signal noise model

International Nuclear Information System (INIS)

Vavrin, J.

1981-01-01

The noise analysis method is discussed in identifying perturbance models and their parameters by a stochastic analysis of the noise model of variables measured on a reactor. The analysis of correlations is made in the frequency region using coherence analysis methods. In identifying an actual specific perturbance, its model should be determined and recognized in a compound model of the perturbance system using the results of observation. The determination of the optimum estimate of the perturbance system model is based on estimates of related spectral densities which are determined from the spectral density matrix of the measured variables. Partial and multiple coherence, partial transfers, the power spectral densities of the input and output variables of the noise model are determined from the related spectral densities. The possibilities of applying the coherence identification methods were tested on a simple case of a simulated stochastic system. Good agreement was found of the initial analytic frequency filters and the transfers identified. (B.S.)

Linear phase formation by noise simulator

International Nuclear Information System (INIS)

Hazi, G.; Por, G.

1998-01-01

A new simulation technique is introduced to study noise propagation in nuclear power plants. Noise processes are considered as time functions, and the dynamic behaviour of the reactor core is modelled by ordinary and partial differential equations. The equations are solved by numerical methods and the results (time series) are considered as virtual measurements. The auto power spectral density and the cross power spectral density of these time series are calculated by traditional techniques. The spectrum obtained is compared with the analytical solution to validate the new simulation approach. After validation, the simulator is expanded to investigate some physical phenomena which are unmanageable by analytical calculations. Propagating disturbances are studied, and the effect of non-flat flux shape on phase curves is demonstrated. Numerical problems also are briefly discussed. (author)

Research and development program in reactor diagnostics and monitoring with neutron noise methods. Stage 7. Final report

International Nuclear Information System (INIS)

Pazsit, I.; Demaziere, C.; Arzhanov, V.

2001-08-01

This report constitutes stage 7 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. A proposal for the continuation of this program in stage 8 is also given at the end of the report. In stage 6, the basic principles of a 3-D fully coupled neutronic/thermal-hydraulic simulator in the frequency domain were presented. The neutronic model relied on the two-group diffusion approximation, whereas the thermal-hydraulic algorithms relied on the so called 'lumped' model. The key element of this simulator was that only the static data were required which could be obtained from the Studsvik Scandpower CASMO-4/TABLES-3/ SIMULATE-3 code package. The simulator was developed with this underlying idea, which means that the calculation of the static fluxes and the eigenvalue were avoided. Depending on what kind of spatial discretization scheme which is used in the noise simulator to calculate the 'leakage' noise, it is not granted that the system remains critical by using the group constants supplied by SIMULATE. Nevertheless, when the system is critical, the balance equations should be fulfilled in all nodes with respect to the discretization scheme used. In concrete terms, the calculation of the static fluxes and eigenvalue can be avoided if the system is brought back to criticality by modifying the cross-sections so that the balance equations are always fulfilled with the chosen spatial discretization scheme. This approach was used in this study with the finite difference scheme. As pointed out in stage 6, the finite difference scheme is relatively inefficient compared to finite elements or nodal methods, but on the other hand it is rather easy to implement. These two more sophisticated schemes are planned to be investigated at a later stage, but for the time being the simulator relying on the finite difference scheme was improved as much as possible so that a 2-D entirely neutronic model could be

BWR core response to fluctuations in coolant flow and pressure, with implications on noise diagnosis and stability monitoring

International Nuclear Information System (INIS)

Blomstrand, J.H.; Andersson, S.A.

1982-01-01

Reactor dynamic tests, utilizing sinuosidal oscillations in pressure and recirculation flow, have been conducted in operating BWRs in Sweden and Finland. Test data recorded, as well as recordings of process noise, have been analyzed in terms of dynamic core properties. The results obtained show good qualitative agreement with model predictions of BWR core dynamics. Model studies can often support interpretation of dynamic information obtained from operating plants. Comparisons between model studies, dynamic tests and process noise may also provide improved understanding of test results and noise patterns; in this way it can be demonstrated that some neutron flux noise is caused by noise in coolant flow and steam flow. From reactor test data nd noise recordings, core stability parameters have been evaluated by a number of methods. These have been found to provide essentially the same results. The cores investigated were found to be very stable under normal operating conditions. In special operating points, outside the normal operating range, higher decay ratios may occur. The experience indicates that for BWR cores, operated at decay ratios above quarter damping, the stability parameters may be identified from the oscillatory behavior of the autocorrelation in the time domain of the neutron flux noise

A random-parametric reactor model with direct feedback and non-white noise

International Nuclear Information System (INIS)

Sako, O.; Taniguchi, A.; Kuroda, Y.

1982-01-01

The effects of multiplicative direct power feedback and non-white reactivity noise on the fluctuations of the neutron density are studied, based on the master equation using the cumulant expansion and the system-size expansion. The results obtained are the following: non-whiteness of reactivity noise reduces the variance of neutron density, as well as the level of the power spectral density. The nonlinear effect of power feedback gives rise to at least a pair of corner frequencies, in contrast to the single corner frequency in linearized case. (author)

Discrimination of ex-core neutron noise signatures using artificial neural networks

International Nuclear Information System (INIS)

Alguindigue, I.E.; Uhrig, R.E.; Cai, M.; Trenty, A.

1993-01-01

The vibratory behavior of the internals in a Pressurized Water Reactor, PWR, can be identified and monitored using ex-core neutron noise data from power detectors located at ionization chambers outside the vessel. The signatures collected from these sensors provide information regarding presence of contacts between the core barrel and the pressure vessel, and more importantly, a means of verifying the integrity of components in the system. This report describes a neural-network-based methodology for identifying the vibration mode of the core barrel, and for detecting a particular family of mechanical failures. Features are extracted from the neutron noise spectra and used for training neural network models to identify the different states of vibratory behavior typically exhibited by PWR'S. The technique was tested on data from twenty eight 900MW pressurized water reactors in France, and the results achieved are over 98% accurate

Implementation of the neutron noise technique under the UBERA-6 project

International Nuclear Information System (INIS)

Gomez, Angel; Bellino, Pablo A.

2009-01-01

Using the neutron noise technique, kinetics parameters estimations and power calibration were performed in the new core of the RA-6 reactor. These activities were carried on under the nuclear start-up of the UBERA-6 project, which consist in the change of core and power increase of the reactor. In a first stage, in joint with the power estimation, the decay constant of the prompt neutrons (α c ) was estimated. Its value was found to agree with the calculations obtained from neutron codes. Lately, in the high power stage, estimators of the calibration factors for the 16 N detection device were obtained. A thorough analysis of the linearity of the instrumentation used was done, and an alternative methodology was applied in order to estimate the aforementioned factor. The calibration factor obtained by the neutron noise technique was in agreement with the one obtained by thermal balance. (author)

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (II). The second experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-01-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the second reactor noise experiment using pseudo random binary sequences (PRBS) was performed on August 30, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 50% of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (III). The third experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-03-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the third reactor noise experiment using pseudo random binary sequences (PRBS) was performed on September 16, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 70% of reactor power and under a normal sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

On-line adaptive line frequency noise cancellation from a nuclear power measuring channel

Directory of Open Access Journals (Sweden)

Qadir Javed

2011-01-01

Full Text Available On-line software for adaptively canceling 50 Hz line frequency noise has been designed and tested at Pakistan Research Reactor 1. Line frequency noise causes much problem in weak signals acquisition. Sometimes this noise is so dominant that original signal is totally corrupted. Although notch filter can be used for eliminating this noise, but if signal of interest is in close vicinity of 50 Hz, then original signal is also attenuated and hence overall performance is degraded. Adaptive noise removal is a technique which could be employed for removing line frequency without degrading the desired signal. In this paper line frequency noise has been eliminated on-line from a nuclear power measuring channel. The adaptive LMS algorithm has been used to cancel 50 Hz noise. The algorithm has been implemented in labVIEW with NI 6024 data acquisition card. The quality of the acquired signal has been improved much as can be seen in experimental results.

The nonlinear treatment in noise analysis and its influence into quality programs

International Nuclear Information System (INIS)

Montesino, Maria Elena; Piedra, Margarita; Rolo, Alberto

2002-01-01

Analysis by noise technique have been utilized in fact for early detection reactor's anomalies, constituting one of the fundamental tools in maintenance condition. There is an international tendency to use this technique in order to guarantee the quality of maintenance control at present. Essentially this technique is an approach from lineal dynamic and insufficient for limit cycle illustrations, which take place after bifurcation point. As non-lineal dynamic (chaos theory) descriptors consider the dynamic characteristics of studied system, their uses guarantee the quality in predictive maintenance by noise. In this paper was analysed the necessity of theory's use as a tool to supplement the analysis by noise technique. (author)

Development of PC-based FFT system for reactor dynamic analysis

International Nuclear Information System (INIS)

Ansari, S.; Baig, A.R.

1993-03-01

A personal computer based fast fourier transform (FFT) analyzer has been developed for frequency spectrum analysis of signals from nuclear reactor. The system can perform window smoothing, computation of auto- and cross-power spectral density, coherence and auto and cross-correlation functions. The feature of 16 analogue signals acquisition with high precision and high sampling frequency makes the analyzer suitable for malfunction diagnosis of nuclear reactors using reactor noise analysis. The development work for the fourier analyzer was undertaken as a part of IAEA research contract no. 5925/RB. The applications of the FFT analyzer are described in reactor transfer function measurements and nuclear instrumentation channels frequency response testing. (author)

Kernel-based noise filtering of neutron detector signals

International Nuclear Information System (INIS)

Park, Moon Ghu; Shin, Ho Cheol; Lee, Eun Ki

2007-01-01

This paper describes recently developed techniques for effective filtering of neutron detector signal noise. In this paper, three kinds of noise filters are proposed and their performance is demonstrated for the estimation of reactivity. The tested filters are based on the unilateral kernel filter, unilateral kernel filter with adaptive bandwidth and bilateral filter to show their effectiveness in edge preservation. Filtering performance is compared with conventional low-pass and wavelet filters. The bilateral filter shows a remarkable improvement compared with unilateral kernel and wavelet filters. The effectiveness and simplicity of the unilateral kernel filter with adaptive bandwidth is also demonstrated by applying it to the reactivity measurement performed during reactor start-up physics tests

Investigation of space-energy effects in the reactivity measurement by neutron noise with excore detectors in a reflected LWR

International Nuclear Information System (INIS)

Lescano, V.H.; Behringer, K.

1982-01-01

Practical application of the zero-crossing correlation method for measuring slightly-subcritical reactivities in a swimming-pool reactor required the use of detector locations in the reflector zone near to the core boundary. Experimental investigations of neutron-noise cross-power spectra showed significant deviations from the point-reactor model at higher frequencies (> 100 Hz). Nevertheless, the use of the point-reactor model was found to be a useful approach in the analysis of the zero-crossing correlation method, yielding results which agreed well with those obtained from the rod-drop method. The theoretical part of the work is concerned with a space-dependent model calculation in two-group diffusion theory to support the experimental findings. The model calculation can explain the trends observed in the neutron-noise spectra as well as the applicability of the point-reactor model to the zero-crossing correlation method. To obtain better insight, the calculations have been extended to neutron-noise spectra when one or both detectors are located in the core zone. In the case of a large core and widely-spaced detectors, with at least one detector in the core zone, a sink frequency appears in the spectra. This effect is well known in coupled-core kinetics. (author)

Determination of local boiling in light water reactors by correlation of the neutron noise; Determination de l'ebullition locale dans les reacteurs a eau legere par correlation du bruit neutronique

Energy Technology Data Exchange (ETDEWEB)

Zwingelstein, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

The power limit of swimming-pool type reactors depends on the phenomenon of the appearance of burn-out. In order to determine this limit we have attempted to detect the local boiling which usually occurs before the burn out. Local boiling has been simulated by an electrically heated plate placed in the core of the reactor Siloette. The study of local boiling, which is based on the properties of the correlation functions for the neutron noise of detectors placed in the core, shows that a privileged frequency occurs in the power spectrum of the noise. It is intended in the future to determine the influence of various parameters on this characteristic frequency. (author) [French] La limitation de la puissance des reacteurs nucleaires de type piscine est due au phenomene d'apparition de 'burn out'. Pour determiner cette limitation, nous nous sommes proposes dans ce rapport de detecter l'ebullition locale qui apparait generalement avant le 'burn out'. L'ebullition locale a ete simulee par une plaque chauffee electriquement et placee dans le coeur du reacteur SILOETTE. L'etude de l'ebullition locale, qui est basee sur les proprietes des fonctions de correlation du bruit neutronique de detecteurs places clans le coeur, fait apparaitre une frequence privilegiee dans le spectre de puissance du bruit. On envisage dans l'avenir, de determiner l'influence des divers parametres sur cette frequence caracteristique. (auteur)

Development of telerobotic manipulators for reactor dismantling work

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

1991-01-01

This paper describes the amphibious electrical manipulators JARM-10, JART-25, JART-100 and JARM-25 which were developed in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. They are multi-functional telerobotic light-duty (10 and 25 daN) and heavy-duty (100 daN) Manipulators which can be used in hostile environments in reactor dismantling work such as high radiation, underwater work and electrical noise. Each manipulator can be operated in either a bilateral master-slave, a teach-and-playback or a programmed control mode. By combining these modes appropriately, it is possible to perform complex tasks of remote handling. The usefulness of the telerobotic systems for dismantling nuclear reactors has been demonstrated by successful application of the JARM-25 for remote underwater dismantlement of highly radioactive reactor internals of complex form of an experimental nuclear power reactor. (author)

Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

International Nuclear Information System (INIS)

Orso, J; Marenzana, A

2012-01-01

Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

Calibration of new I and C at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, Martin; Jurickova, Monika

2011-01-01

The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

Subassembly faults diagnostic of an LMFBR type reactor by the measurement of temperature noise

International Nuclear Information System (INIS)

Kokorev, B.V.; Palkin, I.I.; Turchin, N.M.; Pallagi, D.; Horanyi, S.

1979-09-01

The subassembly faults detection possibility by temperature noise analysis of an LMFBR is described. The paper contains the results of diagnostical examinations obtained on electrically heated NaK test rigs. On the basis of these results the measurement of temperature noise RMS value seems to be a practicable method to detect local blockages in an early phase. (author)

Pattern recognition application for surveillance of abnormal conditions in a nuclear reactor

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Dzwinel, W.

1990-01-01

The system to monitor abnormal conditions in a nuclear reactor, based on the noise analysis of the reactor basic parameters such as power, temperature and coolant flow rate, has been developed. The pattern recognition techniques such as clustering, cluster analysis, feature selection and clusters visualization methods form the basis of the software. Apart from non-hierarchical clustering procedures applied earlier, the hierarchical one is recommended. The system application for IBR-2 Dubna reactor diagnostics is shown. 10 refs.; 6 figs

Investigation of space-energy effects in the reactivity measurement by neutron noise with ex-core detectors in a reflected LWR

International Nuclear Information System (INIS)

Lescano, V.H.; Behringer, K.

1981-11-01

Practical application of the zero-crossing correlation method for measuring slightly subcritical reactivities in a swimming pool reactor required the use of detector locations in the reflector zone near to the core boundary. Experimental investigations of neutron-noise cross-power spectra showed significant deviations from the point reactor model at higher frequencies (> 100 Hz). Nevertheless, the use of the point reactor model was found to be an useful approach in the analysis of the zero-crossing correlation method yielding results which agreed well with those obtained from the rod-drop method. The theoretical part of the work is concerned with a space-dependent model calculation in two-group diffusion theory to support the experimental findings. The model calculation can explain the trends observed in the neutron-noise spectra as well as the applicability of the point reactor model to the zero-crossing correlation method. To obtain better insight, the calculations have been extended to neutron-noise spectra when one or both detectors are located in the core zone. In the case of a large core and widely spaced detectors, with at least one detector in the core zone, a sink frequency appears in the spectra. This effect is well-known in coupled-core kinetics. (Auth.)

Clustering methods and visualization algorithms to aid nuclear reactor operative diagnostics

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Dzwinel, W.

1990-01-01

The software system developed plays the role of the aid to an operator for nuclear reactor diagnostics. The noise analysis of the reactor parameters such as power, temperature and coolant flow rate constitutes the basis of the system. Combination of data acquisition, data preprocessing, clustering and cluster visualization algorithms with heuristic techniques of results analysis, determine the way of its implementation. Two regimes are available. The first one - extended - is recommended for a long term investigations and the second - suppressed for the aid to the reactor operation monitoring. The system has been tested and developed at the JINR IBR-2 pulsed reactor. 13 refs.; 4 figs.; 2 tabs

On-line computer control of a nuclear reactor using optimal control and state estimation methods

International Nuclear Information System (INIS)

Tye, C.

1980-01-01

This paper describes the experimental implementation of a nuclear reactor control system using combined optimal state feedback based on the Quadratic Regulator and state estimation using Kalman filtering techniques. The results obtained from the experiments indicate that a reactor control loop designed using this approach has improved stability margins, greater speed of response and noise filtering properties compared with a conventional reactor control loop. 11 refs

A customized digital monitoring and display system for nonpower reactors

International Nuclear Information System (INIS)

Ficaro, E.P.; Wehe, D.K.

1989-01-01

A digital data acquisition system for monitoring plant variables has been designed and implemented at the University of Michigan's Ford Nuclear Reactor (FNR), a 2-MW open-pool research reactor. The digital data provided by this system will be useful for: improved operator training, real-time experimental calculations, noise analysis, closed-loop control, and expert system applications. This paper describes the analog-to-digital (A/D) transitions and the associated applications and benefits experienced

Determination of the effective delayed neutron fraction in the Coral-I Reactor; Determinacion de la fraccion efectiva de neutrones retardados en el Reactor Coral-1

Energy Technology Data Exchange (ETDEWEB)

Francisco, J L. de; Perez-Navarro, A; Rodriguez-Mayquez, E

1973-07-01

The effective delayed neutron fraction, {beta} eff, has been determined from the measurement of E / {beta}{sup 2}, by means of reactor noise analysis in the time domain, and the neutron detector efficiency, {epsilon}. For the {epsilon} measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm{sup 3}, at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

Noise suppression by noise

OpenAIRE

Vilar, J. M. G. (José M. G.), 1972-; Rubí Capaceti, José Miguel

2001-01-01

We have analyzed the interplay between an externally added noise and the intrinsic noise of systems that relax fast towards a stationary state, and found that increasing the intensity of the external noise can reduce the total noise of the system. We have established a general criterion for the appearance of this phenomenon and discussed two examples in detail.

Reactor condition monitoring and singularity detection via wavelet and use of entropy in Monte Carlo calculation

International Nuclear Information System (INIS)

Kim, Ok Joo

2007-02-01

Wavelet theory was applied to detect the singularity in reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, by wavelet transform after de-noising, singular points can be found easily. To demonstrate this, we generated reactor power signals using a HANARO (a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations and Gaussian noise. We applied wavelet transform decomposition and de-noising procedures to these signals. It was effective to detect the singular events such as sudden reactivity change and abrupt intrinsic property changes. Thus this method could be profitably utilized in a real-time system for automatic event recognition (e.g., reactor condition monitoring). In addition, using the wavelet de-noising concept, variance reduction of Monte Carlo result was tried. To get correct solution in Monte Carlo calculation, small uncertainty is required and it is quite time-consuming on a computer. Instead of long-time calculation in the Monte Carlo code (MCNP), wavelet de-noising can be performed to get small uncertainties. We applied this idea to MCNP results of k eff and fission source. Variance was reduced somewhat while the average value is kept constant. In MCNP criticality calculation, initial guess for the fission distribution is used and it could give contamination to solution. To avoid this situation, sufficient number of initial generations should be discarded, and they are called inactive cycles. Convergence check can give guildeline to determine when we should start the active cycles. Various entropy functions are tried to check the convergence of fission distribution. Some entropy functions reflect the convergence behavior of fission distribution well. Entropy could be a powerful method to determine inactive/active cycles in MCNP calculation

Experience with a PC-based system for noise and DC signal analysis in PWRs

International Nuclear Information System (INIS)

Hashemian, H.M.

1996-01-01

A data acquisition system that was originally developed for noise diagnostics in PWRs was expanded to include DC signal analysis in addition to noise analysis. The system has been used in PWRs for reactor diagnostics, determination of root cause of process anomalies, instrument calibration verification, measurement of drop time of control and shutdown rods, testing of timing and sequencing of control rod drive mechanisms, emergency diesel generator monitoring, etc. These applications are reviewed in this paper. (author)

Nuclear reactor safety protection device

International Nuclear Information System (INIS)

Okido, Fumiyasu; Noguchi, Atomi; Matsumiya, Shoichi; Furusato, Ken-ichiro; Arita, Setsuo.

1994-01-01

The device of the present invention extremely reduces a probability of causing unnecessary scram of a nuclear reactor. That is, four control devices receive signals from each of four sensors and output four trip signals respectively in a quardruplicated control device. Each of the trip signals and each of trip signals via a delay circuit are inputted to a logical sum element. The output of the logical sum circuit is inputted to a decision of majority circuit. The decision of majority circuit controls a scram pilot valve which conducts scram of the reactor by way of a solenoid coils. With such procedures, even if surge noises of a short pulse width are mixed to the sensor signals and short trip signals are outputted, there is no worry that the scram pilot valve is actuated. Accordingly, factors of lowering nuclear plant operation efficiency due to erroneous reactor scram can be reduced. (I.S.)

International symposium on cavitation and multiphase flow noise - 1986

International Nuclear Information System (INIS)

Arndt, R.E.A.; Billet, M.L.; Blake, W.K.

1986-01-01

This book presents the papers given at a symposium on multiphase flow and cavitation. Topics considered at the conference included the development of a cavitation-free sodium pump for a breeder reactor, the stochastic behavior (randomness) of acoustic pressure pulses in the near-subcavitating range, cavitation monitoring of two axial-flow hydroturbines, and noise generated by cavitation in orifice plates with some gaseous effects

Evaluation of signal processing for boiling noise detection. Further analysis of BOR-60 reactor noise data

International Nuclear Information System (INIS)

Ledwidge, T.J.; Black, J.L.

1989-01-01

The present paper deals with investigations of acoustic signals from a boiling experiment performed on the BOR 60 reactor in the USSR. Signals have been analysed in frequency as well as in time domain. Signal characteristics successfully used to detect the boiling process have been found in time domain. A proposal for in-service boiling monitoring by acoustic means is described. (author). 3 refs, 16 figs

Statistical analysis of the vibration loading of the reactor internals and fuel assemblies of reactor units type WWER-440 from deferent projects

International Nuclear Information System (INIS)

Ovcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.; Anikeev, J.; Pljush, A.

2006-01-01

In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed

Experience from and research activities at the Otaniemi TRIGA reactor

International Nuclear Information System (INIS)

Bars, Bruno

1976-01-01

Experience from the Finnish TRIGA Reactor is reported, small changes and improvements in the control console of the Fir-1 reactor have been made. A minicomputer based data collecting system is planned and installed. It will be used for collecting data from operation and radiation monitors including the new isotope laboratory, and also simultaneously smaller experiments such as control rod calibration. A minicomputer is used for on-line reactor noise studies. The automatic uranium analyzer has a maximum sensitivity of 0.03 μg U 235 and 1.2 Th 232 . The system is now used at a sampling rate of about one sample per minute. (author)

Application of the exact distribution pj{sub k} in the determination of kinetic parameters in a reactor; Aplicacion de la distribucion exacta p{sub k} a la determinacion de parametros cineticos de un reactor

Energy Technology Data Exchange (ETDEWEB)

Alca Ruiz, F

1982-07-01

In this report one distribution of neutron counts obtained by a detector placed in a reactor is studied in order to be used in the determination of reactor kinetic parameters such as {beta}/{lambda} and reactivities. The parameters accuracy from this new method is compared with the Feynman and Mogilner method, based too in Reactor Neutron Noise Analysis. These three methods have been applied to JEN-2 reactor and the better accuracy and faster collection of experimental data give some interest to the new method which only requires a good footing code. (Author) 68 refs.

Distributed system for acquisition and analysis of NPP systems noise data

International Nuclear Information System (INIS)

Oprea, F.; Gruia, L.

2001-01-01

Since years '90 the noise analysis for nuclear stations is more frequently used because: - the data can be acquired by making use of the NPP instrumentation; - acquisition may be performed in a steady-state regime, at rated capacity; - reactor dynamical regime data are obtainable although steady-state regime data are used. The signal noise analysis is based on the dynamical information about the reactor operation obtained from instrumentation signal fluctuations as determined in steady-system regime. The small fluctuations are due to the stochastic effects inherent to physical processes such as heat transfer, boiling, turbulences in coolant flow, moderator collective motions, fission process, structural vibrations and pressure oscillations. Specifically, for CANDU type reactors, the results may be applied for: - surveillance of the functioning characteristics of detectors and transducers; - signaling or predicting misfunctioning of detectors and transducers, what implies establishment of optimal time rate for macro-sampling; - signaling or prediction of misfunctioning of reactor core components, what implies setting up methods and procedures for detecting vibrations inside the fuel channels, based mainly on the signals from the ROPT adjacent detectors and on the flow rate transducers. The system of signal acquisition is a distributed, fixed-data acquisition system provided with a limited number of inputs, based on a number of modules of local analog signal processing and conversion to digital representation. The signal, local processing implies noise removal out of signal through continuous current component elimination, an amplification of the signal, low-pass filtering for alias frequency removal, 16 bit analog-to-digital conversion and digital value serial way transmission. The local processing is based on the characteristics of the acquired signal (noise), namely: - the frequency band, 0.01 Hz - 200 Hz; - the signal amplitude level, 16 - 32 MeV superimposed

Application of the exact distribution pjk in the determination of kinetic parameters in a reactor

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1982-01-01

In this report one distribution of neutron counts obtained by a detector placed in a reactor is studied in order to be used in the determination of reactor kinetic parameters such as β/Λ and reactivities. The parameters accuracy from this new method is compared with the Feynman and Mogilner method, based too in Reactor Neutron Noise Analysis. These three methods have been applied to JEN-2 reactor and the better accuracy and faster collection of experimental data give some interest to the new method which only requires a good footing code. (Author) 68 refs

Acoustic Emission for on-line reactor pressure boundary monitoring

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

1985-01-01

The program objective is to develop AE for continuous surveillance to assess flaw growth in reactor pressure boundaries. Technology in the laboratory is being evaluated on structures. Results have demonstrated basic feasibility of the program objective. AE monitoring a long term fatigue test of a pressure vessel demonstrated an instrument system, and the ability to detect unexpected as well as well as known fatigue cracks. Monitoring a nuclear reactor system shows that the coolant flow noise problem is manageable and AE can be detected under simulated operating conditions

Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA{sub 1} and Embalse Argentine NPP; Uso de la tecnica de analisis de ruido para la estimacion de vibraciones mecanicas en las centrales nucleares argentinas Atucha I y Embalse

Energy Technology Data Exchange (ETDEWEB)

Lescano, V.H.; Wentzeis, L.M. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes; Guevara, M.; Moreno, C. [Nucleoelectrica Argentina S.A., Cordoba (Argentina). Central Nuclear Embalse; Pineyro, J. [Nucleoelectrica Argentina S.A., Buenos Aires (Argentina). Central Nuclear Atucha I

1996-07-01

In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

Validation of the dynamics of SDS and RRS flux, flow, pressure and temperature signals using noise analysis technique

International Nuclear Information System (INIS)

Glockler, O.; Cooke, D.F.; Tulett, M.V.

1995-01-01

In 1992, a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro's CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. Critical plant components, instrumentation and processes are monitored on a regular basis, and their dynamic characteristics are verified on-power. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes, reactivity devices and fuel channels induced by moderator/coolant flow, (6) estimating the dynamics and response time of RTD temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (1 0) detecting coolant boiling in fully instrumented fuel channels, (1 1) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the

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

Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

International Nuclear Information System (INIS)

Sackett, J.I.

1983-01-01

This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

Effects of background noise on total noise annoyance

Science.gov (United States)

Willshire, K. F.

1987-01-01

Two experiments were conducted to assess the effects of combined community noise sources on annoyance. The first experiment baseline relationships between annoyance and noise level for three community noise sources (jet aircraft flyovers, traffic and air conditioners) presented individually. Forty eight subjects evaluated the annoyance of each noise source presented at four different noise levels. Results indicated the slope of the linear relationship between annoyance and noise level for the traffic noise was significantly different from that of aircraft and of air conditioner noise, which had equal slopes. The second experiment investigated annoyance response to combined noise sources, with aircraft noise defined as the major noise source and traffic and air conditioner noise as background noise sources. Effects on annoyance of noise level differences between aircraft and background noise for three total noise levels and for both background noise sources were determined. A total of 216 subjects were required to make either total or source specific annoyance judgements, or a combination of the two, for a wide range of combined noise conditions.

Physics of the 252Cf-source-driven noise analysis measurement

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.; Perez, R.B.; Mattingly, J.K.

1997-01-01

The 252 Cf-source-driven noise analysis method is a versatile measurements tool that has been applied to measurements for initial loading of reactors, quality assurance of reactor fuel elements, fuel processing facilities, fuel reprocessing facilities, fuel storage facilities, zero-power testing of reactors, verification of calculational methods, process monitoring, characterization of storage vaults, and nuclear weapons identification. This method's broad range of application is due to the wide variety of time- and frequency domain signatures, each with unique properties, obtained from the measurement. The following parameters are obtained from this measurement: average detector count rates, detector multiplicities, detector autocorrelations, cross-correlation between detectors, detector autopower spectral densities, cross-power spectral densities between detectors, coherences, and ratios of spectral densities. All of these measured parameters can also be calculated using the MCNP-DSP Monte Carlo code. This paper presents a review of the time-domain signatures obtained from this measurement

Regression models of reactor diagnostic signals

International Nuclear Information System (INIS)

Vavrin, J.

1989-01-01

The application is described of an autoregression model as the simplest regression model of diagnostic signals in experimental analysis of diagnostic systems, in in-service monitoring of normal and anomalous conditions and their diagnostics. The method of diagnostics is described using a regression type diagnostic data base and regression spectral diagnostics. The diagnostics is described of neutron noise signals from anomalous modes in the experimental fuel assembly of a reactor. (author)

Power spectral density measurements with 252Cf for a light water moderated research reactor

International Nuclear Information System (INIS)

King, W.T.; Mihalczo, J.T.

1979-01-01

A method of determining the reactivity of far subcritical systems from neutron noise power spectral density measurements with 252 Cf has previously been tested in fast reactor critical assemblies: a mockup of the Fast Flux Test Facility reactor and a uranium metal sphere. Calculations indicated that this measurement was feasible for a pressurized water reactor (PWR). In order to evaluate the ability to perform these measurements with moderated reactors which have long prompt neutron lifetimes, measurements were performed with a small plate-type research reactor whose neutron lifetime (57 microseconds) was about a factor of three longer than that of a PWR and approx. 50% longer than that of a boiling water reactor. The results of the first measurements of power spectral densities with 252 Cf for a water moderated reactor are presented

Calculation of the ex-core neutron noise induced by fuel vibrations in PWRs

International Nuclear Information System (INIS)

Tran Hoai Nam; Cao Van Chung; Hoang Thanh Phi Hung; Hoang Van Khanh

2015-01-01

Calculation of the neutron noise induced by fuel assembly vibrations in two pressurized water reactor (PWR) cores has been performed to investigate the effect of cycle burnup on the properties of the ex-core detector noise. Pendular vibrations of individual fuel assemblies were assumed to occur at different locations in the core. The auto power spectra density (APSD) of the ex-core detector noise was evaluated with the assumption of stochastic vibrations along a random two-dimensional trajectory. The results show that no general monotonic variation of APSD was found. The increase of APSD occurs predominantly for peripheral assemblies. Assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core with the more realistic perturbation model, the effect of the peripheral assemblies will dominate and the increase of the amplitude of the ex-core neutron noise with burnup can be confirmed. (author)

Maximizing noise energy for noise-masking studies.

Science.gov (United States)

Jules Étienne, Cédric; Arleo, Angelo; Allard, Rémy

2017-08-01

Noise-masking experiments are widely used to investigate visual functions. To be useful, noise generally needs to be strong enough to noticeably impair performance, but under some conditions, noise does not impair performance even when its contrast approaches the maximal displayable limit of 100 %. To extend the usefulness of noise-masking paradigms over a wider range of conditions, the present study developed a noise with great masking strength. There are two typical ways of increasing masking strength without exceeding the limited contrast range: use binary noise instead of Gaussian noise or filter out frequencies that are not relevant to the task (i.e., which can be removed without affecting performance). The present study combined these two approaches to further increase masking strength. We show that binarizing the noise after the filtering process substantially increases the energy at frequencies within the pass-band of the filter given equated total contrast ranges. A validation experiment showed that similar performances were obtained using binarized-filtered noise and filtered noise (given equated noise energy at the frequencies within the pass-band) suggesting that the binarization operation, which substantially reduced the contrast range, had no significant impact on performance. We conclude that binarized-filtered noise (and more generally, truncated-filtered noise) can substantially increase the energy of the noise at frequencies within the pass-band. Thus, given a limited contrast range, binarized-filtered noise can display higher energy levels than Gaussian noise and thereby widen the range of conditions over which noise-masking paradigms can be useful.

Reactor Noise: A study of Neutronic Fluctuations in Low-Power Nuclear Reactors, with Special Emphasis on Accurate Time-Domain Analysis. RCN Report

Energy Technology Data Exchange (ETDEWEB)

Dragt, J. B.

1968-10-15

Nuclear reactors can be considered as devices in which nuclear energy is produced as a result of neutron-induced fission reactions. Reactor physics is a branch of applied physics, and is concerned with the physical aspects of the design and study of nuclear reactors. The motivation is the achievement of configurations, which meet certain requirements regarding safety, reliability, economy, etc. The reactor physical method is to study neutron populations in a reactor. This study has two aspects : - the microscopic aspect: a study of the nuclear processes that take place. This aspect belongs to nuclear physics. Reaction probabilities can be expressed in cross sections,which are assumed to be known for the second part: - the macroscopic aspect, concerned with neutron migration and multiplication. All basic features may be traced back to a knowledge of neutron distribution functions. For most phenomena it is sufficient to study the singulet density, i.e. the mean number of neutrons per unit volume, unit velocity, moving in unit solid angle. For the subject of this thesis this singulet density will appear to be insufficient. The theory for the macroscopic aspect is part of statistical mechanics, and is closely related to other statistical theories, for phenomena like transfer of radiation in stellar atmosphere, penetration of radiation in scattering media, cosmic ray showers, etc.

Reactor Noise: A study of Neutronic Fluctuations in Low-Power Nuclear Reactors, with Special Emphasis on Accurate Time-Domain Analysis. RCN Report

International Nuclear Information System (INIS)

Dragt, J.B.

1968-10-01

Nuclear reactors can be considered as devices in which nuclear energy is produced as a result of neutron-induced fission reactions. Reactor physics is a branch of applied physics, and is concerned with the physical aspects of the design and study of nuclear reactors. The motivation is the achievement of configurations, which meet certain requirements regarding safety, reliability, economy, etc. The reactor physical method is to study neutron populations in a reactor. This study has two aspects : - the microscopic aspect: a study of the nuclear processes that take place. This aspect belongs to nuclear physics. Reaction probabilities can be expressed in cross sections,which are assumed to be known for the second part: - the macroscopic aspect, concerned with neutron migration and multiplication. All basic features may be traced back to a knowledge of neutron distribution functions. For most phenomena it is sufficient to study the singulet density, i.e. the mean number of neutrons per unit volume, unit velocity, moving in unit solid angle. For the subject of this thesis this singulet density will appear to be insufficient. The theory for the macroscopic aspect is part of statistical mechanics, and is closely related to other statistical theories, for phenomena like transfer of radiation in stellar atmosphere, penetration of radiation in scattering media, cosmic ray showers, etc

Active Noise Control for Dishwasher noise

Science.gov (United States)

Lee, Nokhaeng; Park, Youngjin

2016-09-01

The dishwasher is a useful home appliance and continually used for automatically washing dishes. It's commonly placed in the kitchen with built-in style for practicality and better use of space. In this environment, people are easily exposed to dishwasher noise, so it is an important issue for the consumers, especially for the people living in open and narrow space. Recently, the sound power levels of the noise are about 40 - 50 dBA. It could be achieved by removal of noise sources and passive means of insulating acoustical path. For more reduction, such a quiet mode with the lower speed of cycle has been introduced, but this deteriorates the washing capacity. Under this background, we propose active noise control for dishwasher noise. It is observed that the noise is propagating mainly from the lower part of the front side. Control speakers are placed in the part for the collocation. Observation part of estimating sound field distribution and control part of generating the anti-noise are designed for active noise control. Simulation result shows proposed active noise control scheme could have a potential application for dishwasher noise reduction.

Determination of the effective delayed neutron fraction in the Coral-I Reactor

International Nuclear Information System (INIS)

Francisco, J. L. de; Perez-Navarro, A.; Rodriguez-Mayquez, E.

1973-01-01

The effective delayed neutron fraction, β eff, has been determined from the measurement of E / β 2 , by means of reactor noise analysis in the time domain, and the neutron detector efficiency, ε. For the ε measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm 3 , at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

Reactor instrumentation and control

International Nuclear Information System (INIS)

Wach, D.; Beraha, D.

1980-01-01

The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

Surveillance of a nuclear reactor by use of a pattern recognition methodology

International Nuclear Information System (INIS)

Dubuisson, B.; Lavison, P.

1980-01-01

A multivariate nonparametric pattern recognition system is described for the surveillance of a high-flux isotope reactor. Two nonparametric methods are worked out: one using the Bayes rule with the Rosenblatt-Parzen estimator for the probability law, and one using the k-nearest neighbor rule. Performances are evaluated by comparing the probability of misclassification between the two chosen classes: the first corresponds to a nonaction of the reactor operator on its power and the second to an action of the pilot. Processing is performed on the power signal of the reactor which is an observation corrupted by noise. The system has been tested on several experiences and implemented to work in real time on the reactor. The aim is to conceive a computer-aided decision system for the reactor's pilot. 17 refs

Monitoring system in reactor dry well

International Nuclear Information System (INIS)

Horie, Akira; Suzuki, Shun-ichi; Yamamoto, Shinji; Kubokawa, Toshihiko; Takagi, Sakae; Yokosawa, Makoto.

1991-01-01

A failed portion of a dry well in a BWR type reactor is monitored and identified from a remote place by a simple structure. That is, laser beams are irradiated under scanning to a portion to be monitored. Then, the reflection light is monitored by a light receiving and monitoring system, and abnormalities such as defects or leaks of monitored portion are optically detected by a remote viewing equipment. With such a constitution, the portion to be monitored in poor operation circumstances of the reactor dry well can always be monitored efficiently from a remote place. The device of the present invention does not undergo the effect of radiation noises, etc. and it is excellent in heat resistance and radiation resistance. (I.S.)

Some properties of zero power neutron noise in a time-varying medium with delayed neutrons

International Nuclear Information System (INIS)

Kitamura, Y.; Pal, L.; Pazsit, I.; Yamamoto, A.; Yamane, Y.

2008-01-01

The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed

A record/replay system for the harmonic analysis of nuclear reactor flux noise

International Nuclear Information System (INIS)

Lawrence, L.A.J.; Corran, E.R.

1960-02-01

The design of a record/replay system for the determination of neutron flux spectra is discussed and circuit details and performance figures are given. Frequency modulation is used with a carrier frequency of 1,000 cycles per second, the complete system having an overall bandwidth of 100 cycles per second. The noise is recorded on magnetic tape and when replayed is analysed into a spectrum by means of a selective amplifier used in conjunction with an infinitely variable speed control on the tape. It is shown that if the lowest spectral frequency is .01 cycles per second a recording time of many hours is necessary. The DIDO noise spectrum is analysed and shown to be contained in a bandwidth of a few cycles per second. (author)

An assessment of noise and vibration measurements taken during the commissioning of Dungeness B

International Nuclear Information System (INIS)

Bickers, B.H.

1983-01-01

The design of the Dungeness B Advanced Gas Cooled Reactor took into account, from an early stage, the vibrational response of structures within the pressure vessel to noise generated by the gas circulators. Theoretical estimates indicated that the circulators, which absorb 13000HP each in circulating the CO 2 coolant at 30 atmospheres around the gas circuit, could produce potentially damaging noise levels. These estimates were backed up with scale model circulator tests and at a later stage with noise transmission tests in the actual reactor vessel and finally by running the circulators in air under atmospheric conditions. The response of the structural components was predicted initially by theoretical methods, such as the Statistical Energy Method, which was used basically as a sorting technique to identify problem areas which could then be tested in acoustic chambers using representative test specimens to give a more reliable estimate of response. Any items of structure for which a significant level of response was predicted, or where representative testing was not possible, were instrumented so that the vibrational response could be measured during the unfuelled pressurized commissioning tests. Some of the more interesting aspects of the results of these tests are discussed together with a comparison of the results with predictions. (author)

Measurements and preliminary interpretation of K-Reactor foundation response to man-made seismic excitation

International Nuclear Information System (INIS)

Lee, R.C.; Stephenson, D.E.

1992-01-01

In support of the Savannah River Technology Center (SRTC) effort to develop K-Reactor seismic design basis ground motions, SRTC monitored local high-explosive tests at a ''free-field'' site adjacent to K-Reactor and on the -40 level on the foundation of K-Reactor. The high-explosive tests were part of the SRTC/United States Geological Survey (USGS) regional refraction and attenuation experiment that used deeply buried high explosive charges near New Ellenton, Snelling, and at more distant South Carolina sites. The primary purpose of the Reactor measurements are to compare the relative amplitude and frequency content of ambient noise and shot generated ground motions measured at the K-Reactor foundation level and in the ''free-field'' so that foundation effects to ground motions can be documented and possibly incorporated in the facility design basis. Data analysis indicates that one of the five high explosive tests provided sufficient excitations at K-Reactor to produce satisfactory signal-to-noise between about 1 Hz and 15 Hz. Within this frequency band, Fourier spectral amplitude ratios of motions recorded within the first 10 seconds of first motion show substantial reductions (30 endash 50%) on shot radial and transverse components for frequencies greater than about 3 to 5 Hz. Approximately 50% reductions between 10 to 15 Hz were seen on vertical component ratios, and amplifications of 100% at 4 Hz and 5 Hz endash 6 Hz

System for unattended surveillance of nuclear reactor behavior

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.

1977-01-01

A multivariate statistical pattern recognition system for reactor noise analysis is presented. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, updating, and dimensionality reduction capabilities. System design emphasizes control of the false-alarm rate. Its abilities to learn normal patterns and to recognize deviations from these patterns were evaluated by experiments at the ORNL High-Flux Isotope Reactor. Power perturbations of less than 0.1% of the mean value in selected frequency ranges were readily detected by the pattern recognition system

Distribution of energy of impulses of the modernized IBR-2 REACTOR

International Nuclear Information System (INIS)

Tayibov, L.A; Mehtiyeva, R.N.; )

2011-01-01

Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

U-233 fuelled low critical mass solution reactor experiment PURNIMA II

International Nuclear Information System (INIS)

Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

1987-01-01

A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

Measurement of two-phase flow variables in a BWR by analysis of in-core neutron detector noise signals

International Nuclear Information System (INIS)

Stekelenburg, A.J.C.; Hagen, T.H.J.J. van der

1996-01-01

In this paper, the state of the art of the measurement of two-phase flow variables in a boiling water reactor (BWR) by analysis of in-core neutron detector noise signals is given. It is concluded that the neutronic processes involved in neutron noise are quite well understood, but that little is known about the density fluctuations in two-phase flow which are the main cause of the neutron noise. For this reason, the neutron noise measurements, like the well known two-detector velocity measurements, are still difficult to interpret. By analyzing neutron noise measurements in a natural circulation cooled BWR, it is illustrated that, once a theory on the density fluctuations is developed, two-phase flow can be monitored with a single in-core detector. (author). 70 refs, 4 figs

Recycling flow rate control device in BWR type reactor

International Nuclear Information System (INIS)

Fujiwara, Tadashi; Koda, Yasushi

1988-01-01

Purpose: To reduce the recycling pump speed if the pressure variation width and the variation ratio in the nuclear reactor exceed predetermined values, to thereby avoid the shutdown of the plant. Constitution: There has been proposed a method of monitoring the neutron flux increase thereby avoiding unnecessary plant shutdown, but it involves a problems of reactor scram depending on the state of the plant and the set values. In view of the above, in the plant using internal pumps put under the thyristor control and having high response to recycling flow rate, the reactor pressure is monitored and the speed of the internal pump is rapidly reduced when the pressure variation width and variation ratio exceed predetermined values to reduce the reactor power and avoid the plant shutdown. This can reduce the possibility of unnecessary power reduction due to neutron flux noises or the possibility of plant shutdown under low power conditions. Further, since the reactor operation can be continued without stopping the recycling pump, the operation upon recovery can be made rapid. (Horiuchi, T.)

Dose rate and total dose dependence of the 1/f noise performance of a GaAs operational amplifier during irradiation

International Nuclear Information System (INIS)

Hiemstra, D.M.

1995-01-01

A pictorial of a sectioned view of the torus of the International Thermonuclear Experimental Reactor (ITER) is shown. Maintenance and inspection of the reactor are required to be performed remotely. This is due to the high gamma radiation environment in vessel during inspection and maintenance activities. The custom GaAs operational amplifier is to be used to readout sensors on the in-vessel manipulator and inspection equipment. The gamma dose rate during maintenance and inspection is anticipated to be 3 Mrad(GaAs)/hour. Here, dose rate and total dose dependence of the 1/f noise performance of a custom GaAs MESFET operational amplifier during irradiation are presented. Dose rate dependent 1/f noise degradation during irradiation is believed to be due to electron trapping in deep levels, enhanced by backgating and shallow traps excited during irradiation. The reduction of this affect with accumulated total dose is believed to be due a reduction of deep level site concentration associated with substitutional oxygen. Post irradiation 1/f noise degradation is also presented.The generation-recombination noise observed post irradiation can be attributed to the production of shallow traps due to ionizing radiation

Apparatus for ultrasonic visualization in sodium (VISUS) and acoustic detection in the Phenix reactor

International Nuclear Information System (INIS)

Lions, M.; Berger, R.; Bret, A.; Buis, H.; Barton, J.

A description is given of two acoustic monitoring systems studied at the Fast Neutron Reactor Department at CEN/Cadarache and used in the Phenix reactor. The first is the active type, and the second, passive. The active apparatus is based on the sonar principle and permits visualizing objects inside a reactor tank, especially the heads of assemblies during handling. The passive apparatus, the acoustic detector, observes the reactor core by analyzing the acoustic noise produced in the reactor core. The electroacoustic converters are in both cases located outside, and the acoustic vibrations are transmitted by wave guides. The two pieces of apparatus can operate in a hostile environment, such as liquid metal at high temperature in the presence of high neutron and gamma fluxes

Practical Ranges of Loudness Levels of Various Types of Environmental Noise, Including Traffic Noise, Aircraft Noise, and Industrial Noise

Directory of Open Access Journals (Sweden)

Sabine A. Janssen

2011-05-01

Full Text Available In environmental noise control one commonly employs the A-weighted sound level as an approximate measure of the effect of noise on people. A measure that is more closely related to direct human perception of noise is the loudness level. At constant A-weighted sound level, the loudness level of a noise signal varies considerably with the shape of the frequency spectrum of the noise signal. In particular the bandwidth of the spectrum has a large effect on the loudness level, due to the effect of critical bands in the human hearing system. The low-frequency content of the spectrum also has an effect on the loudness level. In this note the relation between loudness level and A-weighted sound level is analyzed for various environmental noise spectra, including spectra of traffic noise, aircraft noise, and industrial noise. From loudness levels calculated for these environmental noise spectra, diagrams are constructed that show the relation between loudness level, A‑weighted sound level, and shape of the spectrum. The diagrams show that the upper limits of the loudness level for broadband environmental noise spectra are about 20 to 40 phon higher than the lower limits for narrowband spectra, which correspond to the loudness levels of pure tones. The diagrams are useful for assessing limitations and potential improvements of environmental noise control methods and policy based on A-weighted sound levels.

A real-time BWR [boiling water reactor] stability measurement system

International Nuclear Information System (INIS)

March-Leuba, J.; King, W.T.

1987-01-01

This paper describes the characteristics of a portable, real-time system used for nonperturbational measurements of stability in boiling water reactors. The algorithm used in this system estimates the closed-loop asymptotic decay ratio using only the naturally occurring neutron noise and it is based on the univariate autoregressive methodology

Kinetic parameters of the GUINEVERE reference configuration in VENUS-F reactor obtained from a pile noise experiment using Rossi and Feynman methods

Energy Technology Data Exchange (ETDEWEB)

Geslot, Benoit; Pepino, Alexandra; Blaise, Patrick; Mellier, Frederic [CEA, DEN, DER/SPEx, Cadarache, F-13108 St Paul Lez Durance (France); Lecouey, Jean-Luc [LPC Caen, ENSICAEN, Universite de Caen, CNRS/IN2P3, 6 Bd. Marechal Juin 14050 Caen cedex (France); Carta, Mario [ENEA, UTFISST-REANUC, C.R. Casaccia, S.P.040 via Anguillarese 301, 00123 S. Maria Di Galeria, Roma (Italy); Kochetkov, Anatoly; Vittiglio, Guido [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, BE-2400, Mol (Belgium); Billebaud, Annick [LPSC, CNRS, IN2P3/UJF/INPG, 53 Avenue des Martyrs, 38026 Grenoble cedex (France)

2015-07-01

A pile noise measurement campaign has been conducted by the CEA in the VENUS-F reactor (SCK-CEN, Mol Belgium) in April 2011 in the reference critical configuration of the GUINEVERE experimental program. The experimental setup made it possible to estimate the core kinetic parameters: the prompt neutron decay constant, the delayed neutron fraction and the generation time. A precise assessment of these constants is of prime importance. In particular, the effective delayed neutron fraction is used to normalize and compare calculated reactivities of different subcritical configurations, obtained by modifying either the core layout or the control rods position, with experimental ones deduced from the analysis of measurements. This paper presents results obtained with a CEA-developed time stamping acquisition system. Data were analyzed using Rossi-α and Feynman-α methods. Results were normalized to reactor power using a calibrated fission chamber with a deposit of Np-237. Calculated factors were necessary to the analysis: the Diven factor was computed by the ENEA (Italy) and the power calibration factor by the CNRS/IN2P3/LPC Caen. Results deduced with both methods are consistent with respect to calculated quantities. Recommended values are given by the Rossi-α estimator, that was found to be the most robust. The neutron generation time was found equal to 0.438 ± 0.009 μs and the effective delayed neutron fraction is 765 ± 8 pcm. Discrepancies with the calculated value (722 pcm, calculation from ENEA) are satisfactory: -5.6% for the Rossi-α estimate and -2.7% for the Feynman-α estimate. (authors)

Optimal processor for malfunction detection in operating nuclear reactor

International Nuclear Information System (INIS)

Ciftcioglu, O.

1990-01-01

An optimal processor for diagnosing operational transients in a nuclear reactor is described. Basic design of the processor involves real-time processing of noise signal obtained from a particular in core sensor and the optimality is based on minimum alarm failure in contrast to minimum false alarm criterion from the safe and reliable plant operation viewpoint

Using the coolant temperature noise for measuring the flow rate in the RBMK technological channels

International Nuclear Information System (INIS)

Selivanov, V.M.; Karlov, N.P.; Martynov, A.D.; Prostyakov, V.V.; Lysikov, B.V.; Kuznetsov, B.A.; Pallagi, D.; Khorani, Sh.; Khargitai, T.; Tezher, Sh.

1983-01-01

The problems are considered connected with the possibility of using thermometric correlation method to measure the coolant flow rate in the RBMK reactor technological channels. The main attention is paid to the study of the physical nature of the coolant temperature pulsations and to estimation of the effect of parameters of the primary thermaelectrical converter (TEC) on the results of measurements. In the process of reactor inspections made using the thermometric correlation flowmeter of a special design, the temperature noise distribution in the points of flow rate measurement is studied, the noise intensity and physical nature are determined, as well as the effect of different TEC parameters (TEC inertia and base distance between them) on the measurement accuracy. On the basis of the analysis of the effect on the results of the TEC thermal inertia measured value divergence, tausub(α) and transport time, tau sub(T), a conclusion is made on the necessity of choosing the base distance between TEC with tausub(T)>tausub(d)

Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-01

This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

Methods for solving the stochastic point reactor kinetic equations

International Nuclear Information System (INIS)

Quabili, E.R.; Karasulu, M.

1979-01-01

Two new methods are presented for analysis of the statistical properties of nonlinear outputs of a point reactor to stochastic non-white reactivity inputs. They are Bourret's approximation and logarithmic linearization. The results have been compared with the exact results, previously obtained in the case of Gaussian white reactivity input. It was found that when the reactivity noise has short correlation time, Bourret's approximation should be recommended because it yields results superior to those yielded by logarithmic linearization. When the correlation time is long, Bourret's approximation is not valid, but in that case, if one can assume the reactivity noise to be Gaussian, one may use the logarithmic linearization. (author)

Application of static fuel management codes for determination of the neutron noise using the adiabatic approximation

International Nuclear Information System (INIS)

Garis, N.S.; Karlsson, J.K.H.; Pazsit, I.

2000-01-01

The neutron noise, induced by a rod manoeuvring experiment in a pressurized water reactor, has been calculated by the incore fuel management code SIMULATE. The space- and frequency-dependent noise in the thermal group was calculated through the adiabatic approximation in three dimensions and two-group theory, with the spatial resolution of the nodal model underlying the SIMULATE algorithm. The calculated spatial noise profiles were interpreted on physical terms. They were also compared with model calculations in a 2-D one-group model, where various approximations as well as the full space-dependent response could be calculated. The adiabatic results obtained with SIMULATE can be regarded as reliable for sub-plateau frequencies (below 0.1 Hz). (orig.) [de

Road traffic noise: self-reported noise annoyance versus GIS modelled road traffic noise exposure.

Science.gov (United States)

Birk, Matthias; Ivina, Olga; von Klot, Stephanie; Babisch, Wolfgang; Heinrich, Joachim

2011-11-01

self-reported road traffic noise annoyance is commonly used in epidemiological studies for assessment of potential health effects. Alternatively, some studies have used geographic information system (GIS) modelled exposure to road traffic noise as an objective parameter. The aim of this study was to analyse the association between noise exposure due to neighbouring road traffic and the noise annoyance of adults, taking other determinants into consideration. parents of 951 Munich children from the two German birth cohorts GINIplus and LISAplus reported their annoyance due to road traffic noise at home. GIS modelled road traffic noise exposure (L(den), maximum within a 50 m buffer) from the noise map of the city of Munich was available for all families. GIS-based calculated distance to the closest major road (≥10,000 vehicles per day) and questionnaire based-information about family income, parental education and the type of the street of residence were explored for their potential influence. An ordered logit regression model was applied. The noise levels (L(den)) and the reported noise annoyance were compared with an established exposure-response function. the correlation between noise annoyance and noise exposure (L(den)) was fair (Spearman correlation r(s) = 0.37). The distance to a major road and the type of street were strong predictors for the noise annoyance. The annoyance modelled by the established exposure-response function and that estimated by the ordered logit model were moderately associated (Pearson's correlation r(p) = 0.50). road traffic noise annoyance was associated with GIS modelled neighbouring road traffic noise exposure (L(den)). The distance to a major road and the type of street were additional explanatory factors of the noise annoyance appraisal.

Determination of spatially dependent transfer function of zero power reactor by using pseudo-random incentive

International Nuclear Information System (INIS)

Kostic, Lj.

1973-01-01

Specially constructed fast reactivity oscillator was stimulating the zero power reactor by a stimulus which caused pseudo-random reactivity changes. Measuring system included stochastic oscillator BCR-1 supplied by pseudo-random pulses from noise generator GBS-16, instrumental tape-recorder, system for data acquisition and digital computer ZUSE-Z-23. For measuring the spatially dependent transfer function, reactor response was measured at a number of different positions of stochastic oscillator and ionization chamber. In order to keep the reactor system linear, experiment was limited to small reactivity fluctuations. Experimental results were compared to theoretical ones

Vibration behavior of PWR reactor internals Model experiments and analysis

International Nuclear Information System (INIS)

Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

1975-01-01

In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

Judgments of aircraft noise in a traffic noise background

Science.gov (United States)

Powell, C. A.; Rice, C. G.

1975-01-01

An investigation was conducted to determine subjective response to aircraft noise in different road traffic backgrounds. In addition, two laboratory techniques for presenting the aircraft noise with the background noise were evaluated. For one technique, the background noise was continuous over an entire test session; for the other, the background noise level was changed with each aircraft noise during a session. Subjective response to aircraft noise was found to decrease with increasing background noise level, for a range of typical indoor noise levels. Subjective response was found to be highly correlated with the Noise Pollution Level (NPL) measurement scale.

Identifying functions for ex-core neutron noise analysis

International Nuclear Information System (INIS)

Avila, J.M.; Oliveira, J.C.

1987-01-01

A method of performing the phase analysis of signals arising from neutron detectors placed in the periphery of a pressurized water reactor is proposed. It consists in the definition of several identifying functions, based on the phases of cross power spectral densities corresponding to four ex-core neutron detectors. Each of these functions enhances the appearance of different sources of noise. The method, applied to the ex-core neutron fluctuation analysis of a French PWR, proved to be very useful as it allows quick recognition of various patterns in the power spectral densities. (orig.) [de

Transient plasma estimation: a noise cancelling/identification approach

International Nuclear Information System (INIS)

Candy, J.V.; Casper, T.; Kane, R.

1985-03-01

The application of a noise cancelling technique to extract energy storage information from sensors occurring during fusion reactor experiments on the Tandem Mirror Experiment-Upgrade (TMX-U) at the Lawrence Livermore National Laboratory (LLNL) is examined. We show how this technique can be used to decrease the uncertainty in the corresponding sensor measurements used for diagnostics in both real-time and post-experimental environments. We analyze the performance of algorithm on the sensor data and discuss the various tradeoffs. The algorithm suggested is designed using SIG, an interactive signal processing package developed at LLNL

Early detection of coolant boiling in research reactors with MTR-type fuel

International Nuclear Information System (INIS)

Kozma, R.; Turkcan, E.; Verhoef, J.P.

1992-10-01

A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

Image processing algorithm for robot tracking in reactor vessel

International Nuclear Information System (INIS)

Kim, Tae Won; Choi, Young Soo; Lee, Sung Uk; Jeong, Kyung Min; Kim, Nam Kyun

2011-01-01

In this paper, we proposed an image processing algorithm to find the position of an underwater robot in the reactor vessel. Proposed algorithm is composed of Modified SURF(Speeded Up Robust Feature) based on Mean-Shift and CAMSHIFT(Continuously Adaptive Mean Shift Algorithm) based on color tracking algorithm. Noise filtering using luminosity blend method and color clipping are preprocessed. Initial tracking area for the CAMSHIFT is determined by using modified SURF. And then extracting the contour and corner points in the area of target tracked by CAMSHIFT method. Experiments are performed at the reactor vessel mockup and verified to use in the control of robot by visual tracking

Assessment of the implementation of a neutron measurement system during the commissioning of the Jordan Research and Training Reactor

Energy Technology Data Exchange (ETDEWEB)

Bae, Sang Hoon; Suh, Sang Mun [Division of Research Reactor System Design, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cha, Han Ju [Dept. of Electrical Engineering, Intelligent Power Conversion Laboratory, Chungnam National University, Daejeon (Korea, Republic of)

2017-04-15

The Jordan Research and Training Reactor (JRTR) is the first research reactor in Jordan, the commissioning of which is ongoing. The reactor is a 5-MWth, open-pool type, light-water-moderated, and cooled reactor with a heavy water reflector system. The neutron measurement system (NMS) applied to the JRTR employs a wide-range fission chamber that can cover from source range to power range. A high-sensitivity boron trifluoride counter was added to obtain more accurate measurements of the neutron signals and to calibrate the log power signals; the NMS has a major role in the entire commissioning stage. However, few case studies exist concerning the application of the NMS to a research reactor. This study introduces the features of the NMS and the boron trifluoride counter in the JRTR and shares valuable experiences from lessons learned from the system installation to its early commissioning. In particular, the background noise relative to the signal-to-noise ratio and the NMS signal interlock are elaborated. The results of the count rates with the neutron source and the effects of the discriminator threshold are summarized.

Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.

2012-01-01

Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones

Sodium boiling noise topics in the German Democratic Republic

International Nuclear Information System (INIS)

Ziegenbein, D.

1982-01-01

In German Democratic Republic, the research and development program GDR in the field of nuclear energy is directed' only to selected topics. For instances in the Central Institute for Nuclear Research of the Academy of Sciences a number of tasks related to process diagnosis have been solved as a contribution to the safe and economical operation of our nuclear. power plants. As a result of these investigations noise diagnosis systems have been developed for the primary loops of the 440 MW units. Signals of about 120 detectors can be analysed wth this equipment for plant surveillance and for an early detection of malfunctions. Some topics in the research on Fast Breeder Reactors are directed to selected contributions in the field of process diagnosis. Their solution shall support a fast industrial application of this reactor type. In addition to calculations for reactor core design, primarily related to operational safety of large LMFBRs, noise analysis technique has been applied to acoustic signals for leak detection in sodium heated steam generators as well as for boiling detection in the reactor core. It seems to be promising to investigate whether the same signal analysis methods can be applied to leak and boiling detection, respectively. If this would be possible one could take a standard monitor into consideration for the surveillance of both plant components. Our recent investigations have shown that the beginning of the sodium-water reaction as well as the inspection of sodium boiling is characterized by changes in the statistic signal parameters. Deviations from the normal state can be recognized by measuring actual values of such statistic characteristics of acoustic and/or neutron flux signals. Activities were concentrated mainly on surveillance methods for sodium heated steam generators. A system is in preparation using acoustic as well as chemical methods, taking into account the requirement of diversity for a surveillance system. The boiling

Robust stabilization of burn conditions in subignited fusion reactors using artificial neural networks

International Nuclear Information System (INIS)

Vitela, E. Javier; Martinell, J. Julio

2000-01-01

In this work it is shown that robust burn control in long pulse operations of thermonuclear reactors can be successfully achieved with artificial neural networks. The results reported here correspond to a volume averaged zero-dimensional nonlinear model of a subignited fusion device using the design parameters of the tokamak EDA-ITER group. A Radial Basis Neural Network (RBNN) was trained to provide feedback stabilization at a fixed operating point independently of any particular scaling law that the reactor confinement time may follow. A numerically simulated transient is used to illustrate the stabilization capabilities of the resulting RBNN when the reactor follows an ELMy scaling law corrupted with Gaussian noise. (author)

Feynman-α technique for measurement of detector dead time using a 30 kW tank-in-pool research reactor

International Nuclear Information System (INIS)

Akaho, E.H.K.; Intsiful, J.D.K.; Maakuu, B.T.; Anim-Sampong, S.; Nyarko, B.J.B.

2002-01-01

Reactor noise analysis was carried out for Ghana Research Reactor-1 GHARR-1, a tank-in-pool type reactor using the Feynman-α technique (variance-to-mean method). Measurements made at different detector positions and under subcritical conditions showed that the technique could not be used to determine the prompt decay constant for the reactor which is Be reflected with photo-neutron background. However, for very low dwell times the technique was used to measure the dead time of the detector which compares favourably with the value obtained using the α-conventional method

Feynman-alpha technique for measurement of detector dead time using a 30 kW tank-in-pool research reactor

CERN Document Server

Akaho, E H K; Intsiful, J D K; Maakuu, B T; Nyarko, B J B

2002-01-01

Reactor noise analysis was carried out for Ghana Research Reactor-1 GHARR-1, a tank-in-pool type reactor using the Feynman-alpha technique (variance-to-mean method). Measurements made at different detector positions and under subcritical conditions showed that the technique could not be used to determine the prompt decay constant for the reactor which is Be reflected with photo-neutron background. However, for very low dwell times the technique was used to measure the dead time of the detector which compares favourably with the value obtained using the alpha-conventional method.

Towards a practical Johnson noise thermometer for long-term measurements in harsh environments

International Nuclear Information System (INIS)

Greenen, Adam; Pearce, Jonathan; Cruickshank, David; Bramley, Paul

2015-01-01

The impact of mechanical and chemical changes in conventional sensors such as thermocouples and resistance thermometers can be avoided by instead using temperature sensors based on fundamental thermometry. A prime example of this is Johnson noise thermometry, which is based on measurement of the fluctuations in the voltage of a resistor arising from thermal motion of charge carriers - i.e. the 'Johnson noise'. A Johnson noise thermometer never needs calibration and is insensitive to the condition of the sensor material. It is therefore ideally suited to long-term temperature measurements in harsh environments, such as nuclear reactor coolant circuits, in-pile measurements, nuclear waste management and storage, and severe accident monitoring. There have been a number of previous attempts to develop a Johnson noise thermometer for the nuclear industry, but none have reached commercial exploitation because of technical problems in practical implementation. The main challenge is to extract the tiny Johnson noise signal from ambient electrical noise influences, both from the internal amplification electronics, and from external electrical noise sources. Recent advances in electronics technology and digital signal processing techniques have opened up new possibilities for developing a viable, practical Johnson noise thermometer. We describe a project funded by the UK Technology Strategy Board (now Innovate UK) 'Developing the nuclear supply chain' call, currently underway, to develop a practical Johnson noise thermometer that makes use of innovative electronics for ultralow noise amplification and signal processing. The new electronics technology has the potential to help overcome the problems encountered with previous attempts at constructing a practical Johnson noise thermometer. An outline of the new developments is presented, together with an overview of the current status of the project. (authors)

Towards a practical Johnson noise thermometer for long-term measurements in harsh environments

Energy Technology Data Exchange (ETDEWEB)

Greenen, Adam; Pearce, Jonathan [National Physical Laboratory, Hampton Road, Teddington, TW11 0LW, (United Kingdom); Cruickshank, David; Bramley, Paul [Metrosol Limited, Plum Park Estate, Watling Street, Paulerspury, Northamptonshire, NN12 6LQ, (United Kingdom)

2015-07-01

The impact of mechanical and chemical changes in conventional sensors such as thermocouples and resistance thermometers can be avoided by instead using temperature sensors based on fundamental thermometry. A prime example of this is Johnson noise thermometry, which is based on measurement of the fluctuations in the voltage of a resistor arising from thermal motion of charge carriers - i.e. the 'Johnson noise'. A Johnson noise thermometer never needs calibration and is insensitive to the condition of the sensor material. It is therefore ideally suited to long-term temperature measurements in harsh environments, such as nuclear reactor coolant circuits, in-pile measurements, nuclear waste management and storage, and severe accident monitoring. There have been a number of previous attempts to develop a Johnson noise thermometer for the nuclear industry, but none have reached commercial exploitation because of technical problems in practical implementation. The main challenge is to extract the tiny Johnson noise signal from ambient electrical noise influences, both from the internal amplification electronics, and from external electrical noise sources. Recent advances in electronics technology and digital signal processing techniques have opened up new possibilities for developing a viable, practical Johnson noise thermometer. We describe a project funded by the UK Technology Strategy Board (now Innovate UK) 'Developing the nuclear supply chain' call, currently underway, to develop a practical Johnson noise thermometer that makes use of innovative electronics for ultralow noise amplification and signal processing. The new electronics technology has the potential to help overcome the problems encountered with previous attempts at constructing a practical Johnson noise thermometer. An outline of the new developments is presented, together with an overview of the current status of the project. (authors)

Monitoring the mechanical vibration of in-core detector tubes and fuel channels via ICFD noise analysis

International Nuclear Information System (INIS)

Glockler, O.; Cooke, D.F.; Czuppon, G.J.; Kapoor, K.K.

2000-01-01

Vibrations of core internals are regularly monitored in the CANDU nuclear generating stations of Ontario Power Generation (OPG) via the noise analysis of in-core flux detectors (ICFDs). Voltage signals of standard station instrumentation are recorded by portable multi-channel high-speed high-resolution data acquisition systems, then statistical parameters are derived from the multi-channel time series measurements. Reactor noise analysis is a non-intrusive statistical technique regularly used in system surveillance, diagnostics and in actual operational I and C problems. It utilizes the dynamic information carried by the small fluctuations (noise) of station signals measured around their mean values during steady-state operation. The present paper discusses specific results related to the flow-induced mechanical vibrations of detector tubes and fuel channels. (author)

Direct-reading dial for noise temperature and noise resistance

DEFF Research Database (Denmark)

Diamond, J.M.

1967-01-01

An attenuator arrangement for a noise generator is described. The scheme permits direct reading of both noise resistance and noise temperature¿the latter with a choice of source resistance.......An attenuator arrangement for a noise generator is described. The scheme permits direct reading of both noise resistance and noise temperature¿the latter with a choice of source resistance....

Noise frame duration, masking potency and whiteness of temporal noise.

Science.gov (United States)

Kukkonen, Heljä; Rovamo, Jyrki; Donner, Kristian; Tammikallio, Marja; Raninen, Antti

2002-09-01

Because of the limited contrast range, increasing the duration of the noise frame is often the only option for increasing the masking potency of external, white temporal noise. This, however, reduces the high-frequency cutoff beyond which noise is no longer white. This study was conducted to determine the longest noise frame duration that produces the strongest masking effect and still mimics white noise on the detection of sinusoidal flicker. Contrast energy thresholds (E(th)) were measured for flicker at 1.25 to 20 Hz in strong, purely temporal (spatially uniform), additive, external noise. The masking power of white external noise, characterized by its spectral density at zero frequency N0, increases with the duration of the noise frame. For short noise frame durations, E(th) increased in direct proportion to N0, keeping the nominal signal-to-noise ratio [SNR = (E(th)/N0)(0.5)] constant at threshold. The masking effect thus increased with the duration of the noise frame and the noise mimicked white noise. When noise frame duration and N0 increased further, the nominal SNR at threshold started to decrease, indicating that noise no longer mimicked white noise. The minimum number of noise frames per flicker cycle needed to mimic white noise decreased with increasing flicker frequency from 8.3 at 1.25 Hz to 1.6 at 20 Hz. The critical high-frequency cutoff of detection-limiting temporal noise in terms of noise frames per signal cycle depends on the temporal frequency of the signal. This is opposite to the situation in the spatial domain and must be taken into consideration when temporal signals are masked with temporal noise.

Results and interpretation of noise measurements using in-core self powered neutron detector strings at Unit 2 of the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Gloeckler, O.; Por, G.; Valko, J.

1986-11-01

In-core neutron noise and fuel assembly outlet temperature noise measurements were performed at Unit 2 of Paks Nuclear Power Plant. Characteristics of the reactor and the noise measuring equipment are briefly described. The in-core Rhodium emitter selfpowered neutron detector strings positioned axially above the other show high coherence and linear phase at low frequencies indicating a marked transport effect, not regularly measured in PWRs. The coherence between horizontally placed neutron detectors is small and the phase is zero. A transport effect of different nature is obtained between neutron detectors (in-core and ex-core) and fuel assembly outlet thermocouples. The observed characteristics depend on reactor and fuel assembly power in a way supporting interpretation in terms of coolant density and void content changes and power feedback effects. During routine analysis vibration of 1.1 Hz appeared as a strong peak in the power spectra. The control assembly that was responsible for the observed behaviour could be localized with high certainty. (author)

Identification of multivariate models for noise analysis of nuclear plant

International Nuclear Information System (INIS)

Zwingelstein, G.C.; Upadhyaya, B.R.

1979-01-01

During the normal operation of a pressurized water reactor, neutron noise analysis with multivariate autoregressive procedures in a valuable diagnostic tool to extract dynamic characteristics for incipient failure detection. The first part of the paper will describe in details the equations for estimating the multivariate autoregressive model matrices and the structure of various matrices. The matrices are estimated by solving a set of matrix operations, called Yule-Walker equations. The selection of optimal model order will also be discussed. Once the optimal parameter set is obtained, simple and fast calculations are used to determine the auto power spectral density, cross spectra, coherence function, phase. In addition the spectra may be decomposed into components being contributed from different noise sources. An application using neutron flux data collected on a nuclear plant will illustrate the efficiency of the method

Measurement of reactivity temperature coefficient by noise method in a power reactor

International Nuclear Information System (INIS)

Aguilar, O.

1986-07-01

The temperature reactivity coefficient was estimated on the basis of noise measurements performed in a PWR. The magnitude of the coefficient was evaluated by relating the values of the APSD and CPSD between ex-core neutron detector signals and fuel assembly outlet thermocouple in the low frequency range. Comparison with δρ/δT measurements performed in PWR by standard methods supports the validity of the results. (author)

Biennial activity report of Reactor Engineering Laboratory - 1983 and 1984

International Nuclear Information System (INIS)

Swaminathan, K.; Prahlad, B.

1986-01-01

This report summarises activities of the Reactor Engineering Laboratory for the period January 1983 to December 1984. The report consists of four sections dealing with development of reactor components, prototype tests in sodium, instrumentation development and measurement techniques and noise analysis techniques respectively. As is customary, the activities have been reported in brief but where detailed reports have been prepared the same are referred. The main thrust of the work of the laboratory was still in support of the FBTR which is in an advanced stage of construction and commissioning at Kalpakkam site. Purification of 100 tonnes of commercial grade sodium to reactor grade, pouring of the liquid metal seals and the construction and commissioning of a sodium loop for calibration of the hydrogen leak detector in all represented significant contribution towards FBTR. The section on development of reactor components describes efforts on construction of both electromagnetic and small mechanical sodium pumps. Sodium removal from the control rod drive mechanism by means of vacuum distillation technique has been a useful experience despite some difficulties faced due, possibly, to the presence of extraneous matter in the decontamination set-up. The section on instrumentation development and measurement techniques describes interesting development concerning ultrasonic imaging for under sodium viewing. The last section on noise analysis techniques describes some experience gained in the detection of cavitation in dummy fuel subassembly by means of acoustic technique. The developmental efforts on construction of high temperature acoustic sensors of both piezoelectric and magnetostrictive type have been encouraging. At the end of the report is included a list of technical publications of the laboratory. (author)

The possibilities of the detection of boiling in the reactor core on the basis of acoustic emission method

International Nuclear Information System (INIS)

Liska, J.

1978-01-01

A method is described of detecting the crisis of boiling in the core of PWR type reactors which may lead to fuel element failure. The method can be applied both in reactor development and operation. It is based on boiling detection by acoustic emission testing. The acoustic signal is measured by means of a piezoelectric transducer immersed in water or attached to one end of a waveguide immersed in water. Signals may be measured in either a wide frequency range or in a band approaching the transducer resonance frequency. This is selected such as to be outside the noise band. Experiments in an open water tank and in a water loop showed that under favourable conditions, the acoustic emission testing method was very sensitive, the intensity of acoustic signals was proportional to boiling intensity, and information contained in the emission spectrum shape was primarily of a qualitative nature. The method remains to be tested in an actual reactor where many spurious noise sources exist. (O.K.)

The role of neural networks in reactor diagnostics and control

International Nuclear Information System (INIS)

Pazsit, I.; Kitamura, M.

1997-01-01

Reactor diagnostics and core diagnostics in particular, is an inverse task, just as most other diagnostics. One measures some physical parameter at some position, or the fluctuation thereof, which is given rise by the fluctuation of another parameter, presumably at a different position. In neutron noise diagnostics, the measured quantity is the neutron noise, whereas the cause, fluctuations of the core material, is called the open-quotes noise sourceclose quotes. The relationship between the cause (noise source) and the induced noise (effect) is determined by the physics of the process and can usually be described by a theory. This means that the direct task, calculation of the noise from the noise source, can always be achieved. In diagnostics, however, the process starts from the back end, i.e. one observes the effect of some cause. The task is to infer the cause (noise source) from the effect (induced noise), which is an inverse task (sometimes also called unfolding). The situation is very much similar even regarding control problems. The system state is described by a vector in the parameter space. When it deviates from the desired one, the control parameters need to be changed to bring the system into the desired state. The computation of the change in the state vector, due to a change in the control vector, can be calculated with no difficulty in principle. The control requires however the inverse task to be performed, i.e. to determine which changes of the control vector would bring the state vector into the desired state

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

CERN Document Server

Wang Li

2002-01-01

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

Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

International Nuclear Information System (INIS)

Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

2012-01-01

The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

Active noise control in a duct to cancel broadband noise

Science.gov (United States)

Chen, Kuan-Chun; Chang, Cheng-Yuan; Kuo, Sen M.

2017-09-01

The paper presents cancelling duct noises by using the active noise control (ANC) techniques. We use the single channel feed forward algorithm with feedback neutralization to realize ANC. Several kinds of ducts noises including tonal noises, sweep tonal signals, and white noise had investigated. Experimental results show that the proposed ANC system can cancel these noises in a PVC duct very well. The noise reduction of white noise can be up to 20 dB.

252Cf-source-driven noise analysis measurements for characterization of concrete highly enriched uranium (HEU) storage vaults

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.

1993-01-01

The 252 Cf-source-driven noise analysis method has been used in measurements for subcritical configurations of fissile systems for a variety of applications. Measurements of 25 fissile systems have been performed with a wide variety of materials and configurations. This method has been applied to measurements for (1) initial fuel loading of reactors, (2) quality assurance of reactor fuel elements, (3) fuel preparation facilities, (4) fuel processing facilities, (5) fuel storage facilities, (6) zero-power testing of reactors, and (7) verification of calculational methods for assemblies with the neutron k 252 Cf source and commercially available detectors was feasible and to determine if the measurement could characterize the ability of the concrete to isolate the fissile material

Comparative economics of the breeder and light water reactor

International Nuclear Information System (INIS)

Chow, B.G.

1980-01-01

The issue of breeder timing is studied in this article via a breakeven analysis in which the key driving variables are conveniently segregated into two groups, with uranium price providing the linkage. In one group, the technical and cost characteristics of reactors and fuel cycles determine the uranium breakeven price. In the other group, nuclear demand projections and the uranium supply schedule determine the time paths of uranium price for a given composition of reactor types. The author finds that, even if proliferation risk is ignored, the breeder is not economically competitive with a 30%-improved once-through light water reactor before the year 2030 in the USA and in the world outside communist areas as a whole in 90% of the cases examined. In the exceptional cases, the penalty of delaying commercial breeder introduction to 2030 is small and well within the noise level of long-term energy planning. (author)

Measuring neutron noise induced by travelling air bubbles in a research reactor

International Nuclear Information System (INIS)

Por, G.; Horanyi, S.

1983-05-01

Travelling air bubble experiments carried out in a research reactor confirm an earlier proposed model. The sink structure could be found experimentally in APSD of neutron signals and was used to determine the bubble velocity. The measurements show that neutron detectors measure the velocity of the travelling bubbles, the thermocouples that of the water flow. (author)

The research reactor as a tool in the master in nuclear reactors in Argentina

International Nuclear Information System (INIS)

Notari, Carla

2003-01-01

complete the Master with a seminar: Nuclear Power Plants, and a Thesis. In the frame of the academic plan, multiple activities are organized related to research reactors and also to nuclear power plants. Since the very beginning the performance of selected experiments in a nuclear reactor was recognized as an extraordinary tool to give the students an insight in the principal phenomena associated with the chain reaction and the related engineering problems. This experiments have an intrinsic elevated cost, associated with the relevance of the installation and with the specialized personnel involved. CNEA provides the career with this educational instrument through the Ra-1 and RA-3 reactors located at Constituyentes and Ezeiza Atomic Centers respectively. Various activities are under way but the most established, in the Reactor Physics Course, is the estimation of kinetic parameters in RA-1 reactor. The practice includes three different experiments: Approach to critical and calibration of control rods by the compensation method: Starting in a subcritical state with source the calibration of control rod B1 vs B2 is done by introduction of the first and withdrawal of the second. The methods used are based on the Point Kinetic Model; Measurement of control rods effectivity by the rod-drop method: Separate Rod Drop of rods B1 B2 B3 of the overall ensemble B1 B2 B3 B4 and total scram starting with three withdrawn and one partially inserted, is the procedure followed to estimate the reactivity worth of B1 B2 B3 and scram. The Point Kinetic Model and the Modal Kinetic Model are used; Reactor noise technique for the estimation of reactor parameters: α and Λ. The kinetic parameters are estimated assuring that the Point Kinetic Model is valid (detection chambers near to the core), that the fluctuation of the fission density is the dominant source of the correlated part of neutron noise (measurement at low power, <10kw), the dominance of the fundamental armonic (simultaneous use of

Modeling and diagnostic techniques applicable to the analysis of pressure noise in pressurized water reactors and pressure-sensing systems

International Nuclear Information System (INIS)

Mullens, J.A.; Thie, J.A.

1984-01-01

Pressure noise data from a PWR are interpreted by means of a computer-implemented model. The model's parameters, namely hydraulic impedances and noise sources, are either calculated or deduced from fits to data. Its accuracy is encouraging and raises the possibility of diagnostic assistance for nuclear plant monitoring. A number of specific applications of pressure noise in the primary system of a PWR and in a pressure sensing system are suggested

Prompt neutron decay constant for the Oak Ridge Research Reactor with 20 wt % 235U enriched fuel

International Nuclear Information System (INIS)

Ragan, G.E.; Mihalczo, J.T.

1986-01-01

This paper describes measurements of the prompt neutron decay constant at delayed criticality for the Oak Ridge Research Reactor (ORR) using 20 wt % 235 U enriched fuel and compares these measurements with similar measurements using 93.2 wt % 235 U enriched fuel. This reactor parameter is of interest because it affects the transient behavior of the reactor in prompt criticality accident situations. This experiment is part of a program to investigate the differences in the performance of research reactors fueled with highly enriched and low enriched uranium. The prompt neutron decay constants were obtained using noise analysis measurement techniques for a core with newly fabricated, unirradiated fuel elements

Diagnosis methods based on noise analysis at Cernavoda NPP, Romania

International Nuclear Information System (INIS)

Banica, Constantin; Dobrea, Dumitru

1999-01-01

This paper describes a recent noise analysis of the neutronic signals provided by in-core flux detectors (ICFD) and ion chambers (IC). This analysis is part of on-going program developed for Unit 1 of the Cernavoda NPP, Romania, with the following main objectives: - prediction of detector failures based on pattern recognition; - determination of fast excursions from steady states; - detection of abnormal mechanical vibrations in the reactor core. The introduction presents briefly the reactor, the location of ICFD's and IC's. The second section presents the data acquisition systems and their capabilities. The paper continues with a brief presentation of the numerical methods used for analysis (section 3). The most significant results can be found in section 4, while section 5 concludes about useful information that can be obtained from the neutronic signals at high power steady-state operation. (authors)

MMRW-BOOKS, Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors

International Nuclear Information System (INIS)

Williams, M.M.R.

2007-01-01

Description: Prof. M.M..R Williams has now released three of his legacy books for free distribution: 1 - M.M.R. Williams: The Slowing Down and Thermalization of Neutrons, North-Holland Publishing Company - Amsterdam, 582 pages, 1966. Content: Part I - The Thermal Energy Region: 1. Introduction and Historical Review, 2. The Scattering Kernel, 3. Neutron Thermalization in an Infinite Homogeneous Medium, 4. Neutron Thermalization in Finite Media, 5. The Spatial Dependence of the Energy Spectrum, 6. Reactor Cell Calculations, 7. Synthetic Scattering Kernels. Part II - The Slowing Down Region: 8. Scattering Kernels in the Slowing Down Region, 9. Neutron Slowing Down in an Infinite Homogeneous Medium, 10.Neutron Slowing Down and Diffusion. 2 - M.M.R. Williams: Mathematical Methods in Particle Transport Theory, Butterworths, London, 430 pages, 1971. Content: 1 The General Problem of Particle Transport, 2 The Boltzmann Equation for Gas Atoms and Neutrons, 3 Boundary Conditions, 4 Scattering Kernels, 5 Some Basic Problems in Neutron Transport and Rarefied Gas Dynamics, 6 The Integral Form of the Transport Equation in Plane, Spherical and Cylindrical Geometries, 7 Exact Solutions of Model Problems, 8 Eigenvalue Problems in Transport Theory, 9 Collision Probability Methods, 10 Variational Methods, 11 Polynomial Approximations. 3 - M.M.R. Williams: Random Processes in Nuclear Reactors, Pergamon Press Oxford New York Toronto Sydney, 243 pages, 1974. Content: 1. Historical Survey and General Discussion, 2. Introductory Mathematical Treatment, 3. Applications of the General Theory, 4. Practical Applications of the Probability Distribution, 5. The Langevin Technique, 6. Point Model Power Reactor Noise, 7. The Spatial Variation of Reactor Noise, 8. Random Phenomena in Heterogeneous Reactor Systems, 9. Associated Fluctuation Problems, Appendix: Noise Equivalent Sources. Note to the user: Prof. M.M.R Williams owns the copyright of these books and he authorises the OECD/NEA Data Bank

Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

International Nuclear Information System (INIS)

Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

2011-01-01

Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

Sounds and Noises. A Position Paper on Noise Pollution.

Science.gov (United States)

Chapman, Thomas L.

This position paper focuses on noise pollution and the problems and solutions associated with this form of pollution. The paper is divided into the following five sections: Noise and the Ear, Noise Measurement, III Effects of Noise, Acoustics and Action, and Programs and Activities. The first section identifies noise and sound, the beginnings of…

Behaviour of aged and new flux detectors in Darlington reactors

Energy Technology Data Exchange (ETDEWEB)

Banica, C.; Foster, M., E-mail: Constantin.Banica@OPG.com [Ontario Power Generation, Darlington Nuclear, Bowmanville, Ontario (Canada)

2013-07-01

In-core neutron flux detectors are used for protective and safety functions in the Darlington NGS 'A' CANDU reactors. This paper presents new observations regarding the aging of flux detectors, including response to fuelling, response to unit shutdown and indicators of detector noise. Comparisons of detector signals before and after replacement confirm previous assumptions about aging effects. (author)

Active acoustic leak detection in steam generator units of fast reactors

International Nuclear Information System (INIS)

Oriol, L.; Journeau, Ch.

1996-01-01

Steam generators (SG) of Fast Reactors can be subject to water leakage into the sodium secondary circuit, causing an exothermic chemical reaction with potential serious damage to plant. Within the framework of the European Fast Reactor project, the CEA has developed an active acoustic detection technique which, when used in parallel with passive acoustic detection, will lead to effective leak detection results in terms of reliability and false alarm rates. Whilst the passive method is based on the increase in acoustic noise generated by the reaction, the active method takes advantage of the acoustic attenuation by the hydrogen bubbles produced. The method has been validated: in water, during laboratory testing at the Centre d'Etudes de Cadarache; in sodium, at the ASB loop at Bensberg (Germany) and at AEA Dounreay (Scotland). Full analysis of the tests carried out on the SG of the Prototype Fast Reactor in 1994 during end-of-life testing should lead to reactor validation on the method. (authors)

Non-destructive characterization of the materials for future nuclear reactors

International Nuclear Information System (INIS)

Snopek, J.; Slugen, V.

2016-01-01

For our experiments, we have used Barkhausen noise technique, which is powerful non-destructive method for monitoring stresses in lattices of magnetic materials. We have also used PAS, which is powerful non-destructive method for diagnosing vacancy defects in variable materials. We researched some ODS steels, which are primarily going to be used as fuel cladding or reactor pressure vessel internal components. This thesis describes the behavior of the microstructure of the oxide dispersion strengthened steels at intermediate temperature. Two, in principle, different techniques were used for the characterization of the microstructure of the oxide dispersion strengthened steels thermally aged at 475 grad C and 650 grad C. Both techniques, namely Positron annihilation lifetime spectroscopy (PAS) and Barkhausen noise (BN) measurements are very sensitive to metallurgical modifications and presence of nano-sized imperfections in the crystal lattice. Precipitation of the nano-sized α' phases shift the Barkhausen noise signal. (authors)

A wide-range medition system for TRIGA Mark III Reactor

International Nuclear Information System (INIS)

Vazquez R, M.

1995-01-01

The number of particles emitted by a fission reaction is proportional to the number of fissions generated in the reactor nucleus, what in turn are proportional to the power level of such reactor; this indicates that it is possible could measure the reactor power if the amount of neutrons is measured, or the density of the neutron flux. The power measurement in the reactor is necessary in order to have a control of the same. Several procedures for power measurement exists, one of them is achieved through neutron flux density measurement, that take place in the chain reaction by means of the flow density measurement. The easiest way in order to achieve the neutron flux density measurement is carry out by means of the employment of neutron detectors. To the exit of these detectors an electric pulse taks place every time that a neutron interacts with the sensitive detector part. The work here presented, concrete to the construction of a system of measurement of the nuclear power reactor, that is based on the neutron flux applying some techniques of the neutron noise analysis. (Author)

Continuous surveillance of reactor coolant circuit integrity

International Nuclear Information System (INIS)

1986-01-01

Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

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)

Noise-induced hearing loss: a recreational noise perspective.

Science.gov (United States)

Ivory, Robert; Kane, Rebecca; Diaz, Rodney C

2014-10-01

This review will discuss the real-world risk factors involved in noise-induced hearing loss as a result of common and popular recreational activities prone to mid and high levels of noise exposure. Although there are currently no interventional measures available to reverse or mitigate preexisting hearing loss from noise, we discuss the vital importance of hearing loss prevention from noise exposure avoidance and reduction. Despite a seeming understanding of the effects of noise exposure from various recreational activities and devices, a large percentage of the general public who is at risk of such noise-induced hearing loss still chooses to refrain from using hearing protection instruments. While occupational exposures pose the greatest traditional risk to hearing conservation in selected workers, recreational risk factors for noise-induced hearing loss may be more insidious in overall effect given the indifferent attitude of much of the general public and particularly our youths toward hearing protection during recreational activities. Active counseling regarding the consequences of excessive noise exposure and the potential benefits to hearing from usage of hearing protection instruments is critical to providing best possible care in the hearing health professions.

Determination of the parameters of the point kinetics equation of a nuclear reactor by the quasilinearization technique

International Nuclear Information System (INIS)

Tanomaru, N.

1979-12-01

The problem of parameter identification in a pontual model for a thermal reactor is dealt with using the quasilinearization technique. The model considers one group of delayed neutrons and a heavily non-linear temperature feedback in the reactivity. The parameter prompt neutron generation time and a parameter of the fuel temperatura reactivity coefficient equation are identified simultaneously, considering discrete measurements of the reactor power, during the transient produced by a change in the external reactivity. The influences of the choice of the external reactivity disturbance, of the two parameters values initial guesses, of the interval between measurements and the measurement noise level in the method accuracy and rate of convergence are analysed. For noiseless or low level noise measurements, the method proved to be very effective. (Author) [pt

Activity report of Reactor Physics Division - 1989

International Nuclear Information System (INIS)

1990-01-01

The highlights of the various studies carried out during the year 1989 in Reactor Physics Division are presented in this report in the form of summaries. The topics are organised under the following subjects: (1) nuclear data evaluation, processing and validation, (2) core physics and analysis, (3) reacto r kinetics and safety analysis, (4) noise analysis, and radiation transport and shielding. It is observed that with the restart and operation of FBTR at low power for some time, some of the low power physics experiments were completed and plans and procedures for the remaining physics experiments at intermediate and high power (upto 10 MWt) have been prepared. The lists of publications by the members of Division and the Reactor Physics Seminars held during the year 19 89, are included at the end of the report. (author). refs., figs., tabs

Inspection of Candu Nuclear Reactor Fuel Channels

International Nuclear Information System (INIS)

Baron, J.; Jarvis, G.N.; Dolbey, M.P.; Hayter, D.M.

1986-01-01

The Channel Inspection and Gauging Apparatus of Reactors (CIGAR) is a fully atomated, remotely operated inspection system designed to perform multi-channel, multi-task inspection of CANDU reactor fuel channels. Ultrasonic techniques are used for flaw detection, (with a sensitivity capable of detecting a 0.075 mm deep notch with a signal to noise ratio of 10 dB) and pressure tube wall thickness and diameter measurements. Eddy currrent systems are used to detect the presence of spacers between the coaxial pressure tube and calandria tube, as well as to measure their relative spacing. A servo-accelerometer is used to estimate the sag of the fuel channels. This advanced inspection system was commissioned and declared in service in September 1985. The paper describes the inspection systems themselves and discussed the results achieved to-date. (author) [pt

Support vector machines for nuclear reactor state estimation

Energy Technology Data Exchange (ETDEWEB)

Zavaljevski, N.; Gross, K. C.

2000-02-14

Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.

Support vector machines for nuclear reactor state estimation

International Nuclear Information System (INIS)

Zavaljevski, N.; Gross, K. C.

2000-01-01

Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm

Practical ranges of loudness levels of various types of environmental noise, including traffic noise, aircraft noise, and industrial noise

NARCIS (Netherlands)

Salomons, E.M.; Janssen, S.A.

2011-01-01

In environmental noise control one commonly employs the A-weighted sound level as an approximate measure of the effect of noise on people. A measure that is more closely related to direct human perception of noise is the loudness level. At constant A-weighted sound level, the loudness level of a

Effects of a traffic noise background on judgements of aircraft noise

Science.gov (United States)

Powell, C. A.; Rice, C. G.

1974-01-01

A study was conducted in which subjects judged aircraft noises in the presence of road traffic background noise. Two different techniques for presenting the background noises were evaluated. For one technique, the background noise was continuous over the whole of a test session. For the other, the background noise was changed with each aircraft noise. A range of aircraft noise levels and traffic noise levels were presented to simulate typical indoor levels.

Evaluation of noise pollution in oil extracting region of Lavan and the effect of noise enclosure on noise abatement

Directory of Open Access Journals (Sweden)

F. Golbabaei

2007-09-01

Full Text Available Background and aims   Overexposure to industrial noise pollution induce hearing loss workers. Occupational hearing loss may cause interference whit oral communication, so it may  increase the risk of occupational accidents in workplace as well as affects whit social activities.  This study was conducted on Lavan Island, are of oil extracting regions in the south of Iran. The  object of this study was to evaluate noise pollution and determining the effect of noise enclosure  on noise abatement.   Methods   The noise sources were recognized and noise pressure level was measured by CEL- 440. Noise dose of the exposed workers in high level noise area were measured by CEL 272.   Results   Major noise sources were gas turbines, diesel generators, compressors, fans and gas containing pips, noise contour map revealers that noise level were higher than the recommended national exposure limit. The results of workers noise dose show that their noise exposure were  higher than the recommended value, (p<0.001. Finally, by using the results of noise frequency  analysis of different noise sources, the noise pressure level of each sources was determined in   terms of enclosing them.   Conclusion   By enclosing the noise sources, noise pressure levels can be lowered douse to  acceptable levels but limitation of applying enclosure should be regarded.  

Noise cancellation properties of displacement noise free interferometer

International Nuclear Information System (INIS)

Sato, Shuichi; Kawamura, Seiji; Nishizawa, Atsushi; Chen Yanbei

2010-01-01

We have demonstrated the practical feasibility of a displacement- and frequency-noise-free laser interferometer (DFI) by partially implementing a recently proposed optical configuration using bi-directional Mach-Zehnder interferometers (MZIs). The noise cancellation efficiency was evaluated by comparing the displacement noise spectrum of the MZIs and the DFI, demonstrating up to 50 dB of noise cancellation. In addition, the possible extension of DFI as QND device is explored.

Masking potency and whiteness of noise at various noise check sizes.

Science.gov (United States)

Kukkonen, H; Rovamo, J; Näsänen, R

1995-02-01

The masking effect of spatial noise can be increased by increasing either the rms contrast or check size of noise. In this study, the authors investigated the largest noise check size that still mimics the effect of white noise in grating detection and how it depends on the bandwidth and spatial frequency of a grating. The authors measured contrast energy thresholds, E, for vertical cosine gratings at various spatial frequencies and bandwidths. Gratings were embedded in two-dimensional spatial noise. The side length of the square noise checks was varied in the experiments. The spectral density, N(0,0), of white spatial noise at zero frequency was calculated by multiplying the noise check area by the rms contrast of noise squared. The physical signal-to-noise ratio at threshold [E/N(0,0)]0.5 was initially constant but then started to decrease. The largest noise check that still produced a constant physical signal-to-noise ratio at threshold was directly proportional to the spatial frequency. When expressed as a fraction of grating cycle, the largest noise check size depended only on stimulus bandwidth. The smallest number of noise checks per grating cycle needed to mimic the effect of white noise decreased from 4.2 to 2.6 when the number of grating cycles increased from 1 to 64. Spatial noise can be regarded as white in grating detection if there are at least four square noise checks per grating cycle at all spatial frequencies.

Non-Markovian noise

International Nuclear Information System (INIS)

Fulinski, A.

1994-01-01

The properties of non-Markovian noises with exponentially correlated memory are discussed. Considered are dichotomic noise, white shot noise, Gaussian white noise, and Gaussian colored noise. The stationary correlation functions of the non-Markovian versions of these noises are given by linear combinations of two or three exponential functions (colored noises) or of the δ function and exponential function (white noises). The non-Markovian white noises are well defined only when the kernel of the non-Markovian master equation contains a nonzero admixture of a Markovian term. Approximate equations governing the probability densities for processes driven by such non-Markovian noises are derived, including non-Markovian versions of the Fokker-Planck equation and the telegrapher's equation. As an example, it is shown how the non-Markovian nature changes the behavior of the driven linear process

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)

Noise-induced annoyance from transportation noise: short-term responses to a single noise source in a laboratory.

Science.gov (United States)

Kim, Jaehwan; Lim, Changwoo; Hong, Jiyoung; Lee, Soogab

2010-02-01

An experimental study was performed to compare the annoyances from civil-aircraft noise, military-aircraft noise, railway noise, and road-traffic noise. Two-way within-subjects designs were applied in this research. Fifty-two subjects, who were naive listeners, were given various stimuli with varying levels through a headphone in an anechoic chamber. Regardless of the frequency weighting network, even under the same average energy level, civil-aircraft noise was the most annoying, followed by military-aircraft noise, railway noise, and road-traffic noise. In particular, penalties in the time-averaged, A-weighted sound level (TAL) of about 8, 5, and 5 dB, respectively, were found in the civil-aircraft, military-aircraft, and railway noises. The reason could be clarified through the high-frequency component and the variability in the level. When people were exposed to sounds with the same maximum A-weighted level, a railway bonus of about 3 dB was found. However, transportation noise has been evaluated by the time-averaged A-weighted level in most countries. Therefore, in the present situation, the railway bonus is not acceptable for railway vehicles with diesel-electric engines.

Multi-frequencies ECT algorithms to remove sodium noise in ISI of ferromagnetic SG tubes of FBR

International Nuclear Information System (INIS)

Mihalache, Ovidiu

2012-01-01

The paper presents developments and application of multi-frequency eddy current to be used during In-Service Inspection (ISI) of ferromagnetic steam generator (SG) tubes of Fast Breeder Reactors (FBR). Signal enhancement by means of multi-frequency ECT techniques are validated through 3D simulations of both signals and noise due to sodium forms around SG tube or SP. The purpose of such algorithms is to remove from ECT signal the electromagnetic noise resulting from sodium accumulated outside of SG tubes after SG vessel draining. Finite element method (FEM) simulations are used to analyse different sodium build-up scenarios observed experimentally, and to determine optimal multi-frequency ECT algorithms to suppress the most efficiently sodium noise. Also a new 'window multi-frequency' algorithm is applied and validated using 3-dimensional FEM simulations of SP and sodium forms. (author)

Exact statistical analysis of nonlinear dynamic power reactor models by the Fokker--Planck method. Part II. Reactor with on-off control

International Nuclear Information System (INIS)

Debosscher, A.F.; Dutre, W.L.

1979-01-01

The paper deals with the exact stochastic analysis of the low-frequency neutron density fluctuations in an on-off controlled nuclear power reactor without delayed neutrons and perturbed by Gaussian white reactivity noise. The stochastic process, being Markovian, is completely characterized by its first-order probability density function (pdf) and the transition pdf. The first-order pdf is the normalized solution to the time-independent Fokker--Planck equation (FPE). Using this pdf, a general expression for the moments is obtained. The conditions for stochastic stability in probability, in the mean, and in the mean-square are derived. The time-dependent FPE is solved using the Laplace transform technique, which results in four distinct expressions for the transition pdf, according to the relative magnitude of initial and final reactor power with respect to the regulator level. After Laplace inversion, a physical interpretation of the controller's effect on the stochastic process becomes possible. Finally, making use of the obtained pdf's, the spectral density of the reactor power fluctuations is calculated

Environmental noise and noise modelling-some aspects in Malaysian development

International Nuclear Information System (INIS)

Leong, Mohd Salman; Mohd Shafiek bin Hj Yaacob

1994-01-01

Environmental noise is of growing concern in Malaysia with the increasing awareness of the need for an environmental quality consistent with improved quality of life. While noise is one of the several elements in an Environmental Impact Assessment report, the degree of emphasis in the assessment is not as thorough as other aspects in the EIA study. The measurements, prediction (if at all any), and evaluation tended to be superficial. The paper presents a summary of correct noise descriptors and annoyance assessment parameters appropriate for the evaluation of environmental noise. The paper further highlights current inadequacies in the Environmental Quality Act for noise pollution, and annoyance assessment. Some examples of local noise pollution are presented. A discussion on environmental noise modelling is presented. Examples illustrating environmental noise modelling for a mining operation and a power station are given. It is the authors' recommendation that environmental noise modelling be made mandatory in all EIA studies such that a more definitive assessment could be realised

Annoyance of low frequency noise and traffic noise

DEFF Research Database (Denmark)

Mortensen, F.R.; Poulsen, Torben

2001-01-01

The annoyance of different low frequency noise sources was determined and compared to the annoyance from traffic noise. Twenty-two subjects participated in laboratory listening tests. The sounds were presented by loudspeakers in a listening room and the spectra of the low frequency noises were...

The development of NRTM-turbine flow meter and measurement of the coolant flow rate in-core of 5 MW heating reactor

International Nuclear Information System (INIS)

Zha Meisheng; Wang Xiuqin; Ni Mengchen

1995-01-01

In order to measure the coolant flow rate in-core of 5 MW Heating Reactor the special turbine flowmeter of the type of NRTM has been developed. It consists of a body, a turbine with long screw blade and six pieces of Alnico magnets, and a coil mounted on the body. The advantage of this turbine flowmeter is of low resistance and long working-life. Another advantage is that when the turbine is working or not working its factor of resistance is about the same. It is very important for a natural circulation heating reactor. Because the cable, which is welded to the coil assembly, is long enough to extend out of the reactor vessel to the control room, the signal of flow rate is easy to be disturbed by noise in the case. The traditional method of counting the frequency of the A-C voltage which is induced in the coil has a poor ability for resisting noise. The method of the frequency-spectrum analysis of the frequency of the A-C voltage is used to make sure the accuracy of the measurement of the turbine flow meter. Compared with the method of the count it has a good ability for resisting noise. After three years operation a lot of valuable data were obtained

Steam leak detection in advance reactors via acoustics method

International Nuclear Information System (INIS)

Singh, Raj Kumar; Rao, A. Rama

2011-01-01

Highlights: → Steam leak detection system is developed to detect any leak inside the reactor vault. → The technique uses leak noise frequency spectrum for leak detection. → Testing of system and method to locate the leak is also developed and discussed in present paper. - Abstract: Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.

Evaluating noise abatement measures using strategic noise maps

NARCIS (Netherlands)

Borst, H.C.; Miedema, H.M.E.; Laan, W.P.N. van der; Lohman, W.J.A.

2006-01-01

Noise annoyance due to transportation is widespread in industrialized countries and in urban areas in the developing countries. The European Noise Directive (END) requires an assessment of the noise situation as well as the formulation of action plans for the reduction of the number of people

Noise annoys: effects of noise on breeding great tits depend on personality but not on noise characteristics

NARCIS (Netherlands)

Naguib, M.; Van Oers, K.; Braakhuis, A.; Griffioen, M.; De Goede, P.; Waas, J.R.

2013-01-01

Anthropogenic noise can have serious implications for animals, especially when they communicate acoustically. Yet, the impacts of noise may depend not only on noise characteristics but also on an individual's coping style or personality. We tested whether noise is more disturbing if it masks

Simulation of magnetic hysteresis loops and magnetic Barkhausen noise of α-iron containing nonmagnetic particles

International Nuclear Information System (INIS)

Li, Yi; Li, Qiulin; Liu, Wei; Xu, Ben; Hu, Shenyang; Li, Yulan

2015-01-01

The magnetic hysteresis loops and Barkhausen noise of a single α-iron with nonmagnetic particles are simulated to investigate into the magnetic hardening due to Cu-rich precipitates in irradiated reactor pressure vessel (RPV) steels. Phase field method basing Landau-Lifshitz-Gilbert (LLG) equation is used for this simulation. The results show that the presence of the nonmagnetic particle could result in magnetic hardening by making the nucleation of reversed domains difficult. The coercive field is found to increase, while the intensity of Barkhausen noise voltage is decreased when the nonmagnetic particle is introduced. Simulations demonstrate the impact of nucleation field of reversed domains on the magnetization reversal behavior and the magnetic properties

Genetic algorithm-based neural network for accidents diagnosis of research reactors on FPGA

International Nuclear Information System (INIS)

Ghuname, A.A.A.

2012-01-01

The Nuclear Research Reactors plants are expected to be operated with high levels of reliability, availability and safety. In order to achieve and maintain system stability and assure satisfactory and safe operation, there is increasing demand for automated systems to detect and diagnose such failures. Artificial Neural Networks (ANNs) are one of the most popular solutions because of their parallel structure, high speed, and their ability to give easy solution to complicated problems. The genetic algorithms (GAs) which are search algorithms (optimization techniques), in recent years, have been used to find the optimum construction of a neural network for definite application, as one of the advantages of its usage. Nowadays, Field Programmable Gate Arrays (FPGAs) are being an important implementation method of neural networks due to their high performance and they can easily be made parallel. The VHDL, which stands for VHSIC (Very High Speed Integrated Circuits) Hardware Description Language, have been used to describe the design behaviorally in addition to schematic and other description languages. The description of designs in synthesizable language such as VHDL make them reusable and be implemented in upgradeable systems like the Nuclear Research Reactors plants. In this thesis, the work was carried out through three main parts.In the first part, the Nuclear Research Reactors accident's pattern recognition is tackled within the artificial neural network approach. Such patterns are introduced initially without noise. And, to increase the reliability of such neural network, the noise ratio up to 50% was added for training in order to ensure the recognition of these patterns if it introduced with noise.The second part is concerned with the construction of Artificial Neural Networks (ANNs) using Genetic algorithms (GAs) for the nuclear accidents diagnosis. MATLAB ANNs toolbox and GAs toolbox are employed to optimize an ANN for this purpose. The results obtained show

Sensor signal analysis by neural networks for surveillance in nuclear reactors

International Nuclear Information System (INIS)

Keyvan, S.; Rabelo, L.C.

1992-01-01

The application of neural networks as a tool for reactor diagnostics is examined here. Reactor pump signals utilized in a wear-out monitoring system developed for early detection of the degradation of a pump shaft are analyzed as a semi-benchmark test to study the feasibility of neural networks for monitoring and surveillance in nuclear reactors. The Adaptive Resonance Theory (ART 2 and ART 2-A) paradigm of neural networks is applied in this study. The signals are collected signals as well as generated signals simulating the wear progress. The wear-out monitoring system applies noise analysis techniques, and is capable of distinguishing these signals apart and providing a measure of the progress of the degradation. This paper presents the results of the analysis of these data, and provides an evaluation on the performance of ART 2-A and ART 2 for reactor signal analysis. The selection of ART 2 is due to its desired design principles such as unsupervised learning, stability-plasticity, search-direct access, and the match-reset tradeoffs

Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

Energy Technology Data Exchange (ETDEWEB)

Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)

2012-07-01

Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

The effects of noise reduction technologies on the acceptance of background noise.

Science.gov (United States)

Lowery, Kristy Jones; Plyler, Patrick N

2013-09-01

Directional microphones (D-Mics) and digital noise reduction (DNR) algorithms are used in hearing aids to reduce the negative effects of background noise on performance. Directional microphones attenuate sounds arriving from anywhere other than the front of the listener while DNR attenuates sounds with physical characteristics of noise. Although both noise reduction technologies are currently available in hearing aids, it is unclear if the use of these technologies in isolation or together affects acceptance of noise and/or preference for the end user when used in various types of background noise. The purpose of the research was to determine the effects of D-Mic, DNR, or the combination of D-Mic and DNR on acceptance of noise and preference when listening in various types of background noise. An experimental study in which subjects were exposed to a repeated measures design was utilized. Thirty adult listeners with mild sloping to moderately severe sensorineural hearing loss participated (mean age 67 yr). Acceptable noise levels (ANLs) were obtained using no noise reduction technologies, D-Mic only, DNR only, and the combination of the two technologies (Combo) for three different background noises (single-talker speech, speech-shaped noise, and multitalker babble) for each listener. In addition, preference rankings of the noise reduction technologies were obtained within each background noise (1 = best, 3 = worst). ANL values were significantly better for each noise reduction technology than baseline; and benefit increased significantly from DNR to D-Mic to Combo. Listeners with higher (worse) baseline ANLs received more benefit from noise reduction technologies than listeners with lower (better) baseline ANLs. Neither ANL values nor ANL benefit values were significantly affected by background noise type; however, ANL benefit with D-Mic and Combo was similar when speech-like noise was present while ANL benefit was greatest for Combo when speech spectrum noise was

Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

International Nuclear Information System (INIS)

Blanchard, A.

1999-01-01

'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

Community noise

Science.gov (United States)

Bragdon, C. R.

Airport and community land use planning as they relate to airport noise reduction are discussed. Legislation, community relations, and the physiological effect of airport noise are considered. Noise at the Logan, Los Angeles, and Minneapolis/St. Paul airports is discussed.

The use of boiling noise detection as a protection against faults in sub-assemblies in LMFBRs. Status report of work in the United Kingdom

International Nuclear Information System (INIS)

Burton, E.J.; MacLeod, I.D.

1982-01-01

The development of acoustic techniques for the surveillance of LMFBRs has the objective of providing a monitoring system on-line to give an early warning of incipient failures whilst the reactor is at power at present in the UK. Most attention is being given to safety protection to meet the design proposals for the Commercial Demonstration Fast Reactor (CDFR). One concern in the safety analysis is the hypothetical possibility that a local fault in a subassembly, if undetected could spread to its neighbours, eventually involving the whole core. An early warning of such a potentially propagating event would be given by detecting the boiling of the sodium. The specification of the acoustic technique, and therefore of the development programme, is set by the requirements of the safety analysis and the important features are outlined in the first section of the paper. This is followed by a description of the signal strength from boiling, based on out-of-pile experiments. This signal hat to be discriminated against the background noise arising from thc coolant pumps and the subassembly gag and flow noise. The detection of the acoustic signal may now be made by transducers rather than waveguides provided that the transducers are shown to be reliable enough and the recent work is summarised in the next section. The estimate of the signal/noise ratio depends upon the. transmission of the acoustic waves through the core to the sensor position. There is little experience on transmission in the reactor environment, possibilities for experiments are limited and laboratory tests are being used to improve basic knowledge. Modern computers offer the possibility of improving the sensitivity of detection by advanced data processing and the techniques which are being pursued are briefly described. Although acoustic technology has made great improvements in the last decade, especially in the application of acoustic emission techniques in thermal reactors, there is no experience of the