04/01843 On-line neuro-expert monitoring system for Borssele nuclear power plant: Nabeshima, K. et al. Progress in Nuclear Energy, 2003, 43, (1–4), 397–404

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<ul><li><p>crossover, mutation and random assessment of population for multi- cycle loading pattern (LP) optimization. A detailed description of the chromosomes in the genetic algorithm coded is presented. Artificial Neural Networks (ANNs) have been constructed and trained to accelerate the GA-based search during the optimization process. The whole package, called GAOPT, is linked to the reactor analysis code PANTHER, which performs fresh fuel loading, burn-up and power shaping calculations for each reactor cycle by imposing station-specific safety and operational constraints. GAOPT has been verified by performing a number of tests, which are applied to the Hinkley Point B and Hartlepool reactors. The test results giving loading pattern (LP) scenarios obtained from single and multi-cycle optimization calcu- lations applied to realistic reactor states of the Hartlepool and Hinkley Point B reactors are discussed. The results have shown that the GA/ ANN algorithms developed can help the fuel engineer to optimize loading patterns in an efficient and more profitable way than currently available for multi-cycle refuelling of AGRs. Research leading to parallel GAs applied to LP optimization are outlined, which can be adapted to present day LWR fuel management problems. </p><p>04/01839 Goal-oriented flexible sensing for higher diagnostic performance of nuclear plant instrumentation Takahashi, M. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 105- 111. A framework of goal-oriented sensing has been proposed with the emphasis on the integration of an inference mechanism with an active sensing module, which is equipped with a driving mechanism and a set of sensors. The validity of the dynamic failure identification based on the proposed framework of goal-oriented sensing has been examined under realistic experimental conditions using a small-scale test loop. Though the loops are small in scale, the validity of introducing mobile sensing mechanism has been successfully shown through the exper- iments emulating realistic failure conditions. </p><p>04/01840 Identification method of stochastic nonlinear dynamics using dynamical phase analysis-application to Forsmark data Watanabe, F. et al. Annals of Nuclear Energy, 2004, 31, (4), 375-397. A new method is proposed to identify stochastic non-linear dynamics of neutron fluctuation with 'dynamical phase analysis (DPA)'. Availability of the method DPA is demonstrated with the use of neutron noise data from Forsmark. Also, the features of competing spatial global and regional modes in Forsmark data are discussed physically from the identified stochastic non-linear model. </p><p>0401841 Liquid velocity in upward and downward air- water flows Sun, X. et al. Annals of Nuclear Energy, 2004, 3I, (4), 357-373. Local characteristics of the liquid phase in upward and downward air- water two-phase flows were experimentally investigated in a 50.8-mm inner-diameter round pipe. An integral laser Doppler anemometry (LDA) system was used to measure the axial liquid velocity and its fluctuations. No effect of the flow direction on the liquid velocity radial profile was observed in single-phase liquid benchmark experiments. Local multi-sensor conductivity probes were used to measure the radial profiles of the bubble velocity and the void fraction. The measurement results in the upward and downward two-phase flows are compared and discussed. The results in the downward flow demonstrated that the presence of the bubbles tended to flatten the liquid velocity radial profile, and the maximum liquid velocity could occur off the pipe centreline, in particular at relatively low flow rates. However, the maximum liquid velocity always occurred at the pipe centre in the upward flow. Also, noticeable turbulence enhancement due to the bubbles in the two-phase flows was observed in the current experimental flow conditions. Furthermore, the distribution parameter and the void-weighted area-averaged drift velocity were obtained based on the definitions. </p><p>04/01842 Measurement of the Syrian MNSR delayed neutron fraction and neutron generation time by noise analysis Khamis, I. et al. Annals of Nuclear Energy, 2004, 31, (3), 331-341. Delayed neutron fraction /3 and prompt neutron generation time were determined for the Miniature Neutron Source Reactor of Syria using noise analysis technique. Small reactivity perturbations, step-wise and impulse in time, were introduced into the reactor at low power level i.e. zero-power. Power and reactivity versus time were obtained. Using the generalized least square algorithm and transfer function analysis, measurement of both the delayed neutron fraction and the neutron generation time were made. The MNSR values obtained for the prompt neutron generation time and delayed neutron fraction are 78.3 + 1.3 gs and 7.94 4- 0.11 x 10 -3 respectively. Both measured values of fl and ), were found to be very consistent with previously measured and calculated ones reported in the Safety Analysis Report. </p><p>05 Nuclear fuels (scientific, technical) </p><p>04101843 On-line neuro-expert monitoring system for Borssele Nuclear Power Plant Nabeshima, K. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 397- 404. A new method for an on-line monitoring system for the nuclear power plants has been developed utilizing the neural networks and the expert system. The integration of them is expected to enhance a substantial potential of the functionality as operators support. The recurrent neural network and the feed-forward neural network with adaptive learning are selected for the plant modeling and anomaly detection because of the high capability of modeling for dynamic behavior. The expert system is used as a decision agent, which works on the information space of both the neural networks and the human operators. The information of other sensory signals is also fed to the expert system, together with the outputs that the neural networks generate from the measured plant signals. The expert system can treat almost all known correlation between plant status patterns and operation modes as a priori set of rules. From the off-line test at Borssele Nuclear Power Plant (PWR 480 MWe) in the Netherlands, it was shown that the neuro-expert system successfully monitored the plant status. The expert system worked satisfactorily in diagnosing the system status by using the outputs of the neural networks and a priori knowledge base from the PWR simulator. The electric power coefficient is simultaneously monitored from the measured reactive and active electric power signals. </p><p>04/01844 Preliminary measurements of the prompt neutron decay constant in MASURCA Rugama, Y. et al. Progress in Nuclear Energy, 2003, 43, (1~), 421-428. Pulse counting techniques have been used to measure the prompt decay constant c~= (/3 - p)/)~ in the MASURCA reactor of CEA at critical state. The data has been analysed in time domain using Rossi-c~ and Feynman-c~ techniques, and in frequency domain using the cross power spectral density. The Rossi-c~ technique has been studied using one and two detectors. Due to the strong inherent spontaneous fission source, the one-detector variant gives a very strong white-noise signal, which is absent in the two-detector method. Because each neutron detected recorded not only a pulse, but also an echo after 120 ns, corrections had to be made to the theory applied. The Feynman-c~ technique is even more sensitive to the echo in the signals, and quite large corrections had to be made. Nevertheless the results obtained are in reasonable agreement with those of the correlation methods. For both measurement techniques, experiments of long duration are needed to get accurate results. The results obtained agree within 10% with calculations. The prompt decay constant has also been measured with a continuous current technique. From the cross power spectral density thus obtained, the s-value is in agreement with that of the pulse counting techniques. </p><p>04/01845 Reactor noise measurements in the safety and regulating systems of CANDU stations Gl6kler, O. Progress in Nuclear Energy, 2003, 43, (1-4), 75-82. Reactor noise measurements of safety and regulating system instru- mentation are performed in the CANDU nuclear power stations of Ontario Power Generation (OPG) and Bruce Power. Station signals included in the noise measurements are in-core flux detectors (ICFD), ion chambers (I/C), flow transmitters, pressure transmitters, and resistance temperature detectors (RTD). Their frequency dependent noise signatures are regularly measured during steady-state operation, and are used for parameter estimation and anomaly detection. The specific applications include the following areas: (1) Flux noise measurements to detect and characterize (a) anomalies of in-core flux detectors, ion chambers and their electronics, (b) mechanical vibration of fuel channels and in-core detector tubes induced by coolant/ moderator flow. (2) Pressure and flow noise measurements to estimate the in-situ response times of flow/pressure transmitters and their sensing lines installed in the reactor's coolant loops. (3) Temperature noise measurements to estimate the in-situ response times of thermal- well or strap-on type RTDs installed in the reactor's coolant and moderator loops. </p><p>04/01846 Software integration for monitoring systems with high flexibility Suzudo, T. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 405 411. A new scheme of software integration methodology is applied to the implementation of a neural-network-based real-time monitoring aid system. In this scheme, the data communication between modules is established by means of connecting a standard I/O stream of a module to that of another. The methodology enables easy coupling of software modules with the least modifications of existing source code, and it can make distributed software systems highly flexible, portable and testable. </p><p>Fuel and Energy Abstracts July 2004 257 </p></li></ul>