05 Nuclear fuels (scientific, technical)
05 NUCLEAR FUELS
Scientific, technical
04/01830 A PHWR with slightly enriched uranium. About the first core Notari, C. Annals of Nuclear Energy, 2004, 31, (3), 303-309. Many different studies have been performed in Argentina regarding the use of slightly enriched uranium in pressurized heavy water reactor (PHWR) nuclear plants. These referred mainly to operating plants so that a transition had to be considered from the current natural uranium fuel cycle to the slightly enriched one. The first argentine NPP, Atucha I, commissioned in 1974 and designed to operate with natural uranium was converted to slightly enriched uranium in a process that was initiated in 1995 and culminated in 2001. In this analysis, where new PHWR plants are considered, technical and economical arguments are presented which favour the use of a natural uranium initial core. The levelized fuel costs are shown to be practically insensitive to the first core and a fast transition between the two cycles is more influential than an initial enriched core.
04/01831 AMTEC/TE static converters for high energy utilization, small nuclear power plants EI-Genk, M. S. and Tournier, J.-M. P. Energy Conversion and Management, 2004, 45, (4), 511 535. A conceptual design of static converters for small, co-generation modular nuclear power plants is developed and analysed. Each converter is comprised of an alkali metal thermal-to-electric conversion (AMTEC) top cycle and thermoelectric (TE) bottom cycle cooled by natural convection of air. In addition to electricity production at a net efficiency in the low thirties, the small nuclear power plants with AMTEC/TE converters could provide co-generation heat for space heating, seawater desalination and/or high temperature process heat or steam. For the potassium AMTEC/TE converter, the conversion efficiency is about 1% point higher, the hot side temperature >100 K lower and the co-generation heat is slightly lower than for the sodium AMTEC/TE converter when operated at the same anode vapour pressure. On the other hand, when operated at the same hot side temperature, the efficiency of electricity production of power plants with K-AMTEC/TE converters could be ~25% higher, while the co- generation thermal energy for space heating is ~25% lower than with Na-AMTEC/TE converters. The present analysis showed that K- and Na-AMTEC/TE converters could be sized to produce up to 64 and 81 kW~ each at hot side temperatures of 1030 and 1150 K, respectively, while achieving >90% total utilization of the nuclear reactor's fission energy for the plant.
04/01832 Analysis of pressure signals using a Singular System Analysis (SSA) methodology Palomo, M. J. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 329- 336. Pressure signals are used for control and monitoring the safety in nuclear power plants. These pressure transmitters are tested period- ically. In particular, on;y the transmitters checked to analyse the response times are of interest. The noise analysis is a very efficient tool to obtain adequate response times, but although this method can produce accurate results if performed properly, inherent problems in the tests can produce invalid results, for example if the test signal is oscillatory. An approach to remove the oscillatory contamination contribution from the pressure transmitter signal is the Singular Spectrum Analysis (SSA). This method is based on the fact that the dynamics associated with the signal can be described statistically in a linear way by its principal axes. Many or most directions in the embedding space can be associated with noise and typically a small number of these principal directions can explain the main dynamic characteristics of the signal; these principal directions are related with the singular values of the correlation matrix signal. This methodology has been applied successfully to several pressure signals of a typical Pressurized Water Reactor, which show an oscillatory contribution of 0.6 Hz. The SSA method removes this oscillatory contamination very efficiently.
04/01833 Assessment of linear and non-linear autoregressive methods for BWR stability monitoring Manera, A. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 321-327. A benchmark has been performed to compare the performances of exponential autoregressive (ExpAR) models against linear autoregres- sive (AR) models with respect to boiling water reactor stability monitoring. The well-known March-Leuba reduced-order model is used to generate the time-series to be analysed, since this model is able
256 Fuel and Energy Abstracts July 2004
to reproduce the most significant non-linear behaviour of boiling water reactors (i.e. converging, diverging and limit-cycle oscillations). In this way the stability characteristics of the signals to be analysed are known a priori. An application to experimental time-traces measured on a thermalhydraulic natural circulation loop is reported as well. All methods perform equally well in determining the stability character- istics of the analysed signals.
04•01834 Data reconciliation and fault detection by means of plant-wide mass and energy balances Sunde, S. and Berg, 0 . Progress in Nuclear Energy, 2003, 43, (1-4), 97- 104. A plant-wide mass and heat balance model was fitted to 39 values of temperature, pressure and flow for a turbine cycle supplied by steam from a BWR. The calculated results were in good agreement with the measurements. For example, deviations between calculated and measured feedwater flow were usually in the order of 2 kg/s. The quality of the fitted results may be expressed as an overall assesment index Q stating directly the probability that the process is free of faults. The results are discussed with respect to the inclusion versus exclusion of the turbines' flow passing characteristics in the plant model.
04/01835 Development of a BWR control rod pattern design system based on fuzzy logic and knowledge Francois, J.-L. et al. Annals of Nuclear Energy, 2004, 31, (4), 343-356. The development of a system for the control rod pattern design of boiling water reactors based on fuzzy logic and knowledge is presented in this paper. The fuzzy logic module makes use of membership functions and fuzzy rules to control the neutron multiplication factor and the maximum relative nodal power by acting over the control rods' relative movement and the in-core water flow. The Mamdani implication process is used to evaluate the membership functions. The multiplication factor is controlled in order to bring the reactor to the critical state, and the maximum relative nodal power is controlled to maintain the main thermal limits in the core under their design values. Knowledge rules are implemented to govern the control rods' movement and the coolant flow in the core. The fuzzy logic control system is linked to the three-dimensional neutronic and thermal- hydraulic steady state simulator CM-PRESTO. The system was tested with the cycle 10 of Laguna Verde Nuclear Power Plant, Unit 1. The results obtained show that a very reasonable control rod programming can be achieved with a quite simple methodology. The obtained cycle length is comparable with that achieved with a Haling based simulation.
04/01836 Development of microphone leak detection technology on Fugen NPP Shimanskiy, S. et al. Progress in Nuclear Energy, 2003, 43, (1 4), 357 364. A method of leak detection, based on high-temperature resistant microphones, was originally developed in JNC to detect leakages with flow rates from 1 m3/h to 500 m3/h. The development performed on Fugen is focused on detection of a small leakage at an early stage. Specifically, for the inlet feeder pipes the leak rate of 0.2 gpm (0.046
s m / h ) has been chosen as a target detection capability. Evaluation of detection sensitivity was carried out in order to check the capability of the method to satisfy this requirement. The possibility of detecting and locating a small leakage has been demonstrated through the research.
04/01837 Dynamics modeling and stability analysis of a fluidized bed nuclear reactor Lathouwers, D. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 437- 443. A theoretical model describing the coupling of neutronics, thermo- hydraulics and fluidization in a fluidized bed nuclear reactor is presented. The stability of the system is investigated by linearizing and perturbing the system around its equilibrium points and identifying the root loci of the system. It is found that within the operational range, the eigenvalues are located in the negative part of the phase plane, implying linear stability. Simulations of transient conditions are performed, viz. a hypothetical start-up transient and a quasi-static transient related to noise resulting from stochastic movements of the fuel particles. These simulations show that although the total power of the reactor may reach high values, the fuel temperature is well below safety limits at all times.
04/01838 Genetic algorithms and artificial neural networks for loading pattern optimisation of advanced gas-cooled reactors Ziver, A. K. et al. Annals of Nuclear Energy, 2004, 31, (4), 431-457. A non-generational genetic algorithm (GA) has been developed for fuel management optimization of Advanced Gas-Cooled Reactors, which are operated by British Energy and produce around 20% of the UK's electricity requirements. An evolutionary search is coded using the genetic operators; namely selection by tournament, two-point
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.
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.
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.
04•01841 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.
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.
05 Nuclear fuels (scientific, technical)
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.
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.
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.
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.
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