BWR stability boundary surveillance

Neutron noise analysis can make great contributions in order to prevent the power instability event during the reactor starting up process. There is not reason to suppose that the stability boundary in the operation map, cannot drift away getting closer to the real operation point. Noise analysis can perform a double eye boundary drift surveillance by fitting the noise time series to an autoregressive model, and calculating the complex and the real pole. The complex pole accounts for the Decay Ratio and the real pole confirms this accounting. The real pole has a hidden relation with the chaotic behavior of the power, and it is related with the bubble residence time. In case of instability, the real pole vanishes.     

Neutron signal characterization of the power instability event at Laguna Verde BWR

On January 24, 1995, a power instability event occurred in Laguna Verde, a BWR/5 commercial plant. Recorded power oscillations were studied from the point of view of noise analysis. The 723-s long recorded signal comes from average power range monitors and was bad conditioned for noise analysis practice; it was neither stationary in mean, nor in variance. The signal first stage corresponds to the stable reactor; the third stage, to the unstable reactor. There was a second intermediate stage regarded as a transition one. The signal was preconditioned and divided in small blocks. Noise was analysed within each block in the amplitude, frequency and time domains. The analysis was aimed at on early recognition of instability by using the noise to discriminate between stable, transition and unstable state, regardless of the domain chosen for analysis. The experience obtained from studying real events, not depending on any physical model, are the ground for making safer operation procedures.     

Results of acoustic and temperature noise measurements on sodium cooled electrically heated 7 rod bundle

Assembly cooling deficiency in a LMFBR is one of the most important safety problems for reactor design and operation.

Studies on early detection and diagnosis of local accident by means of noise analysis techniques have been initiated at CNEN. Acoustic and temperature noise measurements have been carried out on a 7 rod bundle during slow power transients up to boiling conditions. The test section, simulating the italian PEC reactor fuel element, was mounted on ENA-2 sodium loop located at the CSN Casaccia.

Acoustic noise spectral analysis up to 32 kHz shows the appearance, in presence of boiling, of power increase at certain frequencies. Power spectra and rms values are updated and recorded every 0.3 sec and show large variations going from single phase to boiling.

Temperature noise spectral analysis shows that the power, between 1 and 50 Hz, increases, in presence of boiling, by a factor bigger than 30. It has been tested the sensitivity of other indicators of the temperature fluctuations, like skewness and flatness, to reveal boiling.     

核电站主蒸汽隔离阀气流诱发振动与噪声的数值分析

核电站主蒸汽隔离阀技术要求严、安全系数高,是核电装备国产化进程中的重点攻关项目之一。某些核电站在运行过程中其主蒸汽隔离阀出现过气流诱发振动与噪声,最终导致阀门部件磨损的现象。本文通过数值方法对主蒸汽隔离阀内三维湍流流场进行模拟分析,研究找出产生振动与噪声的流场诱因。本工作的数值分析结果与现场的测量结果相符。     

Measurement and analysis of noise spectra – an effective diagnostics method for nuclear reactors

The basic principles of measuring and analyzing nuclear-reactor noise are described. The results obtained for the IBR-2 reactor by the noise method are presented. It is shown that analysis of the noise spectra of the power and the main reactor parameters makes it possible to find deviations from normal reactor operation at the level 10^–6 ∆k/ k of the change in reactivity.     

Investigation of Resonant Power Oscillation in Halden Boiling Water Reactor Based on Noise Analysis

In the Halden Boiling Water Reactor (HBWR), a resonant power oscillation with a period around 0.04 Hz is observed at power levels higher than about 9.5 MWt. Although this resonant oscillation is not so strong as to affect the normal reactor operation, it is significant, from the viewpoint of reactor diagnosis, to reveal the cause of this oscillation as well as to understand its characteristics.

Noise analysis based on the autoregressive (AR) modeling technique together with spectral and correlation analyses is performed to investigate the driving source, which indicates that it is attributed to the dynamic interference with the reactor of heat exchange process in two parallel-connected steam transformers.

The present study demonstrates the effectiveness of the technique applied here for determining the so-called noise source inducing variations of quantities in a system together with its applicability to various problems in the field of reactor noise analysis and diagnosis.     

基于功率密度谱的压水堆核电厂中子噪声特性研究

分析了压水堆核电厂中子噪声功率密度谱的计算方法,利用该方法以核电厂堆内构件振动监测系统长期的监测数据为基础,计算了中子噪声的功率密度谱,分别分析了百万千万级核电厂、不同功率核电厂和不同燃料周期核电厂中子噪声功率密度谱特性。结果表明,通过分析压水堆核电厂的中子噪声功率密度谱特性,能有效的认识压水堆核电厂堆内构件的振动行为,为压水堆核电厂堆内构件状态分析提供了基础。      

Real time reactor noise diagnostics for the Borssele (PWR) nuclear power plant

After the upgrade of Borssele NPP in 1997, core cycle 24, the power plant operated three years more with 91% availability. The authority of the power plant decided to enhance and upgrade the reactor trend monitoring and plant information recording system with higher frequencies than the plant data processing system (PPS) as well as installing a flexible and multiple-purpose reactor noise analysis system which may support the reactor maintenance group with on-line and off-line capabilities for several different signal processing applications. Two measuring and monitoring systems were built in 2001 and fully taken in implementation during the start-up of the new core 28. In this sense, the new system was used in power operation during the 29^th of September 2001. This paper will introduce the measuring system, the operational tasks, and the results obtained so far on the real-time core-barrel motions (CBM) and the two-primary coolant pump vibrations measured through the reactor noise analysis.     

Bubbles in the sensing line of nuclear power plant pressure transmitters: the shift of spectrum resonances

Noise analysis is a valuable tool when applied to predictive maintenance problems in nuclear power plants. When focused to the capacitive pressure sensors it can detect the presence of undesirable bubbles in the sensing line. From the noise signal power spectrum density a resonant peak shifting towards the lower frequency region, due to bubbles appearance, is observed. In this work an explanation for the peak displacement is given; besides, a method to estimate the void fraction from the shift is developed. Both results benefit the on-line maintenance of pressure sensors without disturbing plant operation.     

Numerical tools applied to power reactor noise analysis

In order to be able to calculate the space- and frequency-dependent neutron noise in real inhomogeneous systems in two-group theory, a code was developed for the calculation of the Green's function (dynamic transfer function) of such systems. This paper reports on the development as well as the test and application of the numerical tools employed. The code that was developed yields the space-dependence of the fluctuations of the neutron flux induced by fluctuating properties of the medium in the two-group diffusion approximation and in a two-dimensional representation of heterogeneous systems, for both critical systems and non-critical systems with an external source. Some applications of these tools to power reactor noise analysis are then described, including the unfolding of the parameters of the noise source from the induced neutron noise, measured at a few discrete locations throughout the core. Other concrete applications concern the study of the space-dependence of the Decay Ratio in Boiling Water Reactors, the noise-based estimation of the Moderator Temperature Coefficient of reactivity in Pressurized Water Reactors, the modeling of the beam- and shell-mode core-barrel vibrations in Pressurized Water Reactors, and the investigation of the validity of the point-kinetic approximation in subcritical systems driven by an external source. In most of these applications, calculations performed using the code are compared with at-power plant measurements. Power reactor noise analysis applications of the above type, i.e. core monitoring without disturbing plant operation, is of particular interest in the framework of the extensive program of power uprates worldwide.     

CMOS low-noise switched charge sensitive preamplifier for CdTe andCdZnTe X-ray detectors

CdTe and CdZnTe X-ray detector arrays for imaging and spectroscopy provide low capacitance current sources with low leakage currents. The optimal shaping time for low-noise operation is relatively high in CMOS analog channels that provide the readout for these detectors. The shaper is centered at lower frequencies, and thus the 1/f noise from the electronics is the main noise source that limits the resolution of the channel. The optimal dimensions of the input stage MOSFET are determined by this noise. In this paper a design criterion for the optimization of the resolution and the power consumption in a 1/f noise dominated readout is introduced. A readout based on CMOS switched charge sensitive preamplifier without feedback resistor has been designed and fabricated in the CMOS 2-μ low-noise analog process provided by MOSIS. This design provides high sensitivity and the possibility to integrate a large number of channels with low power consumption. Measurements of the performance of a first prototype chip are presented     

γ射线相机噪声功率谱分析

本文从理论上分析了γ射线相机噪声的主要来源,并在能量分别为0.2 MeV和1.25 MeV γ射线源上标定不同辐照剂量条件下的噪声功率谱。研究结果表明,噪声主要由MCP像增强器增益起伏引起,在空间频率约为0.633 1 mm^-1处相对强度较强,该峰值在0.2 MeV和1.25 MeV γ射线辐照条件下基本保持不变,噪声功率谱为圆形对称分布,其强度随空间频率增大而快速减小。     

噪声法测量临界装置温度系数实验研究

为了用噪声法测量核电站反应堆的负慢化剂反应性温度系数(MTC),本文在铀溶液临界装置上研究噪声法实验测量反应性温度系数αT,并与周期法测量的αT进行比较。结果表明,两种方法测量的αT趋势基本一致。由于铀溶液临界装置中溶液的反应性温度效应与核电站反应堆的慢化剂温度效应的机理相似,因此本文利用噪声法测量铀溶液临界装置的αT对于核电站反应堆利用噪声法测量MTC有一定的参考价值。     

非线性动力学理论在反应堆噪声分析中的应用研究

利用非线性动力学理论，给出了反应堆噪声功率振荡的非线性特性的差别方法，并对实际信号进行了处理。结果表明热态模拟运行时，反应堆的功率振荡易发生Hopf分叉，圆周形吸引子是存在的，随机噪声会对维数产生较大的干扰，但基本未改变吸引子的形状。该方法对反应堆的详细分析有一定的实际意义。     

Main sources of background acoustic noise of the main circulation loop in VVÉR-1000 reactors

Conclusions are drawn, on the basis of experimental data obtained during experimental operation of the acoustic system for monitoring the leaks in the No. 3 unit of the Kalinin nuclear power plant, concerning the source of background noise which competes with the noise due to the efflux of coolant through leaks in the first loop. Two main sources are identified. The first one at pressure less than 8 MPa in the loop is due to pulsations of bubbles of undissolved gas in the coolant. The second source, which produces correlated noise in all circuits of the loop, is due to noise originating in the deaerator of the purge-makeup system. The high degree of correlation of the intensity of background noises is important for filtering such noise. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 161–166, September, 2007.     

The use of the noise analysis for the detection of the pellet- cladding mechanical interaction

The pellet-cladding mechanical interaction (PCMI) has been classically studied with the cladding elongation signal versus the rod power, but a different method can be used to evaluate the PCMI. It is based on the noise analysis techniques and uses the coherence function between the neutron flux and the cladding elongation noise signals detecting PCMI. This technique has been studied in the Halden experimental reactor. The mean coherence versus average linear heat rate shows two different regions that can be fitted with two straight lines, with the crossing point between them being taken as the threshold for PCMI. In addition, noise analysis in the time domain allows us to measure an elongation response time to a power step. This result can be applied to power instability events in BWRs.     

Development of Reactor Noise Monitor

A reactor noise monitor has been developed to provide a tool that serves the real time routine function of detecting anomalous states of operation in power reactors. In this monitor, RMS (root mean square) values and average frequencies of noise signals are calculated on a real time basis. The average frequency is defined as the frequency averaged by weighting with PSD (power spectral density) of the noise signal and proved to be equivalent to the ratio of an RMS value of the differentiated noise signal to an RMS value of the noise signal itself. Thus, the average frequency can be obtained on a real time basis by using a differentiator, an RMS calculator and a divider. The frequency components of 12 kinds of signals ranging of 0.01–36 Hz are divided into two or three frequency regions by band-pass filters and the RMS value and the average frequency of each frequency region are obtained by the digital calculation method. The performance of this monitor was tested by using a simulated anomalous signal with a peak on its PSD. From the test, it is proved that this monitor can detect more than changes of the RMS value and average frequency. The functions of this monitor were also confirmed at the actual BWR power plant.

In addition, a method to detect peak location of PSD is developed and the accuracy of the peak frequency measurement of the proposed method is investigated to be satisfactory.     

Vibration- and pressure signals as sources of information for an on-line vibration monitoring system in PWR power plants

Higher demands with regard to the safety and reliability of reactor primary components require methods to get an idea of the mechanical state of the plant at any time during operation and to recognize failures already in their developing phase. Reactor vibration monitoring systems are being developed which are based on the analysis of vibration signals, neutron noise and pressure fluctuation signals. The special role vibration and pressure signals can play in such a system is investigated by the analysis of extensive preoperational tests at different PWRs. The theoretical foundation for the application of these signals to vibration monitoring are developed in the special case of the Stade nuclear power plant. The pressure vessel of this reactor performs pendular and verical vibrations. They are excited mainly by pressure fluctuations generated by the coolant flow, by standing waves, or by the revolution of the coolant pumps.

For interpreting the spectra measured during the preoperational test and during power operation and for clearing up changes of these spectra, which will signalise incipient failures, model investigations are of predominant importance. Two mechanical models, a pendular and a vertical one, simulate the two kinds of vibration sufficiently which can be seen in comparing the calculated frequency response with the measured vibrations.     

核电站中跌落零件冲击信号的时频综合报警方法

在对核电站中跌落零件冲击信号和复杂背景噪声进行时域分析和周围图法功率谱估计的荐,提出了跌落零件时频综合报警方法。该方法首先将背景噪声带通滤波,然后对信号时域预报警及频域报警确认。     

Method of Experimental Power Reactor Noise Analysis without Use of High Power Reactor

A method is proposed for conducting power reactor noise analysis without recourse to an actual high power reactor. The basic concept is to simulate the power reactor noise by integrating the different elements constituting the actual reactor noise, such as the random noise-generating force, the zero-power reactor transfer function, and feed- back loops between the reactor power and the noise-generating force.

For the simulation study, a nucleate boiling noise generator and a single feed-back loop were divised to permit experiments with flux-related heater input using the fast neutron source reactor YAYOI of the University of Tokyo.

This report discusses problems encountered in applying the proposed method to the simulation of power reactor noise, and presents some of the results obtained: Two kinds of change in amplitude were observed in the normalized auto power spectral density of the neutron flux, emanating form: (a) fluctuating displacements of the boiling zone bottom boundary, and (b) fluctuating number of passing vapor voids.

Significant differences in the resulting data were observed between the runs performed with constant and with flux-related heater input in the case of step response experiment, but not in steady noise analyses. This apparent ineffectiveness of the feed-back system is due to the relatively small value of the product of the reactivity power coefficient and the reactor power (–5×10-²%Δk/k) in the present feed-back experiment.