中子噪声在核反应堆吊篮振动监测中的应用研究

通过对临界装置堆芯吊篮激励振动引发中子噪声实验得到的功率谱密度(PSD)进行分析,证实了从中子噪声PSD中获得吊篮振动特性(各阶特征频率)是可行的,并给出了中子噪声探测器PSD幅度与吊篮振动幅度之间的比例因子(刻度因子)的计算方法.针对临界装置测量获得的中子噪声PSD和吊篮振动PSD,实际计算了对应吊篮各阶振型的刻度因子.本文证实,可以通过中子噪声分析,给出吊篮结构的振动频率和振动位移,证实了中子噪声在堆内构件振动监测领域的有效性.     

基于中子噪声分析的某核电厂堆芯吊篮梁型振动特征研究

研究了基于堆外电离室中子噪声信号监测压水堆核电厂反应堆吊篮的方法,通过计算电离室中子噪声的互功率密度谱、相干和相位,分析得到了堆芯吊篮梁型振动的频率;利用该方法,计算获得了某正常运行状态下压水堆核电厂换料周期内堆芯吊篮梁型振动频率和中子噪声功率谱幅度的变化趋势,结果说明了在反应堆正常运行状态下,随着堆芯燃耗的增加,吊篮梁型振动频率发生了微小漂移,频率变小,该频率处中子噪声功率谱幅度变大。     

反应堆堆内部件振动的中子噪声物理模型

介绍了一种用来描述堆内部件振动引发中子噪声的物理模型。本模型基于控制论的观战将反应堆动态方程与随机振动方程结合起来,并在振动方程描述中考虑了非线性项。通过在零功率堆的实验和分析,证明了该模型能够描述堆内部件振动引发中子噪声的动态特性,为反应堆运行时堆内部件振动特性的判别提供了一种手段。     

基于实验数据的吊篮表面脉动压力分析

基于堆内构件缩比模型流致振动实验实测得到的吊篮表面脉动压力数据,分析了吊篮表面不同位置脉动压力功率谱密度的分布特征,并对脉动压力功率谱密度的相关性进行分析得到相关长度的特性。结果表明,吊篮表面的脉动压力功率谱密度随频率的增大快速减小然后趋于平缓,是一种频率成份十分丰富的宽带衰减谱,在吊篮同一高度区域的脉动压力功率谱密度基本相同,不同高度区域的脉动压力功率谱密度的能量差别较大;脉动压力功率谱密度的相关长度随频率递增而急剧减小然后趋于常值;吊篮流致振动响应对脉动压力功率谱密度的影响较小,将吊篮流致振动简化为弱耦合问题是合理的。      

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。      

基于中子噪声分析技术的核电站振动监测系统研制

以秦山核电二期扩建工程松脱部件与振动监测系统(KIR)供货项目为背景,研制出了VMS C1201堆内构件振动监测系统.该系统由4个加速度通道和8个中子噪声通道组成,采用PXI总线技术以及虚拟仪器、数据库管理和监测报告自动生成技术,信号调理采用现场可编程门阵列(FPGA)程控技术,各通道信号采用同步处理技术;监测软件采用原始数据存储,并提供开放式接口.该系统具有时程分析、自谱与互谱分析以及压力容器、吊篮和燃料组件振动监测功能.     

核反应堆吊篮振动引发中子噪声的实验研究

利用临界装置,开展了吊篮振动引发中子噪声实验研究.在反应堆稳定运行工况下,采取随机信号、窄带扫频和单频3种激励方式激励吊篮振动.在最大激励力下,吊篮最大振幅21μm,主要振幅在3～10μm.对中子噪声频谱分析显示,本底中子噪声信号频谱幅度随频率近似呈指数衰减.对临界装置的吊篮,振动引发的中子噪声频谱主要分布在数十Hz以上,容易分辨.研究中采用了高通数字滤波技术,改善了对特征频率的观察和识别.随机力激励吊篮,激发出了吊篮和围板的低、中频振动引发的中子噪声特征谱;窄带扫频激励吊篮,在20Hz～70Hz范围的扫频窄带内,中子噪声频谱主要出现70Hz以上的特征谱线,在70Hz～110Hz范围内,主要出现低于扫频窄带频率的特征谱线,在110～180Hz范围内,主要出现125Hz左右的特征谱线,从180～190Hz起的窄带扫频中,只能激励125Hz左右的特征谱线;30Hz～150Hz频率范围内单频激励吊篮,主要出现几个特征谱峰.由此鉴别出吊篮振动模态频率为76.2、125.0、181、215、239.7Hz.     

脉冲堆瞬发负温度系数的噪声测量法

根据反应堆堆内中子信号和温度信号的相关性，以及中子动力学方程和热力学绝热方程，利用噪声信号分析技术，在频域内建立了测量脉冲堆堆芯瞬发负温度系数的物理模型，同时，利用反应堆稳态运行是中子信号以及温度信号和扰动信息，通过自回归滑动和ＡＲＭＡ模型，得到中子噪声信号以温度噪声信号的功率谱密度（ＰＳＤ）再在频域内，通过最优化方法得到脉冲堆堆芯瞬发负温度系数值，该值与理论值相符。     

基于反应堆噪声技术的零功率堆绝对功率测量系统设计

系统利用反应堆噪声分析技术测量零功率堆缓发临界状态下的堆动态参数和绝对功率.在靠近堆芯对称布置两个γ补偿电离室,电离室探测到的反应堆中子噪声信号经测量系统调理和采集后,运用LabVIEW平台开发的软件对噪声信号进行分析处理得到互谱密度,用非线性最小二乘法拟合得到动态参数,并由算法计算出零功率堆的绝对功率.经实地测量,动态参数和绝对功率与堆运行参数相吻合.     

基于压力容器振动信号分析的堆芯吊篮壳型振动特性研究

核反应堆堆芯吊篮的振动状态直接关系到堆芯的安全运行,但堆芯吊篮处于高温和强辐照环境下,无法直接在吊篮上布置传感器测量其振动。本文利用安装在压力容器上的加速度计间接监测吊篮的振动,通过对多核电机组压力容器振动信号相干谱、自功率谱和互功率谱进行分析,获得吊篮壳型振动频率和振幅,并将分析结果与秦山核电厂二期1号机组试验实测值进行比较,分析结果与试验结果相近。研究表明通过对压力容器振动信号的监测与分析,能够有效识别堆芯吊篮壳型振动特性,为吊篮状态评价提供基础。      

Heats of Vaporization of UO2 and UO from UO2

In an earlier paper, a stochastic model of a power reactor has been proposed by the present author on the premise that the coolant-flow through a core is usually accompanied by random variations in the flow-rate, which are eventually largely responsible for the internal reactivity fluctuations.

In the present work, this model is extended to three different reactor systems: (a) where there exists a relaxation process corresponding to the effect of buoyant flow; (b) where a control or fuel element vibrates randomly, due to coolant flow-rate fluctuations; (c) where there are fluctuations in the inlet temperature with a non-white spectrum.

The noise spectra are derived for various state quantities with use made of the Langevin procedure. The theory is illustrated by referring chiefly to the neutron noise spectra, and comparing with the results of observations.

It is shown that the noise sources in question contribute significantly to the spectra, as compared with a low frequency component due to an inherent noise source in the coolant flow. In particular, a strong resonance peak of the spectra arises from the coupling between the random mechanical vibrations and coolant the flow-rate fluctuations.     

Power Noise Spectra of a Water Reactor in Low Frequency Region

The power noise measurements were carried out on the Kyoto University Reactor at various reactor power levels under natural convection for two kinds of core configuration with different temperature coefficients of reactivity. Analysis of the results revealed strong noise in the low frequency region at higher power levels, even with a core configuration of essentially zero temperature coefficient of total reactivity.

A conventional theoretical model assuming a white noise input in reactivity and a reactor transfer function with temperature-reactivity feedbacks was adopted for comparison with the observed data, but it resulted in considerable disagreement at higher power levels.

As a result, it is concluded that there must exist other power effects not yet clarified in the reactivity feedbacks or in the noise inputs to the reactor, which are responsible for the experimental noise spectrum distortion in the low frequency region.     

Reactor Noise Analysis of Swimming Pool Type Reactor

Reactor power frequency spectrum measurements at various power levels (0.2 W, 1 W, 5 W, 100 W, 500 W, 5 kW and 100 kW) were made with HTR** (swimming-pool type). A low frequency AC amplifier, a magnetic tape recorder, a frequency selective amplifier with twin-T filters, a multiplier, and an integrator were used. Speed-up and speed-down techniques of tape recorder were convenient for extending the frequency range of the analysis.

The measured frequency spectrum of reactor noise determined the modulus of zero power transfer function, and indicated a prompt neutron mean life time of (7.58±1.58) × 10^-5 sec based on an effective delayed neutron fraction of 0.0082. The calculations were made with a HIPAC-103 computer. At higher power, some resonance peaks were found in the low frequency region. The absolute value of reactor power obtained by noise analysis agreed within 3% with the power meter indication at the power below 5kW.     

Identification of Transfer Functions in Feedback System and Inverse Problem

A unified theory of identification in a feedback system is presented from the viewpoint of inverse problem on reactor noise. A power reactor described by a 4 block feedback model is expressed in terms of equivalent correlation functions as a 5 block feedback model with an additional loop, if one of subsystems of the reactor becomes unstable, however, the total system is stable by negative feedback effects, and if noise sources are not independent. The addition of a loop in the feedback system leads errors in estimation of open loop transfer functions. In the case, modified open loop transfer functions are identified instead of original ones. From the viewpoint of reactor diagnosis, notes on system identification are discussed on an open loop transfer function in a low power reactor and on effects of diagonalization in covariance space of noise sources.     

Some Relations of Neutronic Noise with Fluctuation of Inlet Coolant Temperature and Vibration of a Control Rod Obtained by Simultaneous Measurements at KUR

Experiments on reactor noise were conducted at KUR. Depending on the operating condition of the reactor, the cause of the noise are classified into the following four types.

1. Zero-power noise source due to the branching process of fission neutrons and/or due to random bombardment of neutrons to the detector—under natural circulation of coolant and at essentially zero-power level.

2. Coolant temperature fluctuation due to natural convection—under natural circulation and at relatively high power level.

3. Flow induced vibration of shim control rods—under forced circulation of coolant and at low power level.

4. Fluctuation of inlet coolant temperature—under forced circulation and near the maximum power level.

Vibration of a spare shim control rod and fluctuation of inlet coolant temperature were measured simultaneously with the neutronic noise. Then the noise sources of the types (3) and (4) were verified. The vibration of a control rod has a broad spectrum in low frequency region besides the large peak at 14 Hz. The fluctuation of inlet coolant temperature is non-white noise and consists of large low frequency component. The theoretically predicted sink structures in the neutronic PSD relating to the transit time of inlet coolant temperature fluctuation through the core were not observed in the experimental results.     

A New Method for Low Sensitivity Design of a Controller for a Load Following BWR Power Plant

A new method has been proposed in which the controller is approximated by two low order models, the second being the sensitivity model of the first. Thus a Kind of piecewise solution is proposed in which the system retains the same order irrespective of the number of parameters to be considered for low sensitivity design. The usefulness of the proposed technique is illustrated in the controller design for a direct cycle BWR power plant of 457 MW(thermal) with recirculation control. The mathematical model includes the reactor kinetics, hydrodynamics of the recirculation loop, pressure transients, and the load frequency control system. The response of system variables such as frequency, neutron power, and reactor pressure are plotted with the low sensitivitty controller. The sensitivity function of frequency has been plotted using the conventional and proposed low sensitivity controller for 20 per cent variation in parameter values. The method is specifically recommended for controller design of large size systems.     

Identification of reactor internals'' vibration modes of a Korean standard PWR using structural modeling and neutron noise analysis

The vibration characteristics of a Korean standard PWR reactor internals have been estimated through a three-dimensional finite element analyses and verified by using the mode separated power spectral density functions obtained from the ex-core neutron noise signals. Also the natural vibration modes of the fuel assembly have been identified measuring both the ex-core and the in-core neutron noise signals which are close to each other. As a result, the fundamental bending mode frequency of the reactor internal structure is found to be around 8 Hz and the fundamental shell mode frequency 14.5 Hz, respectively. It is also shown that the fundamental bending mode frequency of the fuel assembly is 2.3 Hz and the 2^nd bending mode frequency 5.8 Hz, respectively. These results can be used for the supplements of the Korean standard PWR's CVAP (Comprehensive Vibration Assessment Program) data.     

Measurement of Temperature Fluctuations in a Natural Convection Water Loop

Investigations of reactor noise in water-cooled research reactors show that the power spectral density rises in the low frequency domain. The cause of this phenomenon is often attributed to fluctuations in the coolant temperature, but this has never been proved experimentally. The present experiment is an attempt in this direction.

The temperature fluctuation in a natural convection heat transfer loop decoupled from neutronics was measured and analyzed in the frequency domain. The test section of the loop had a rectangular channel measuring 5 mm × 50 mm in cross section and 500 mm in length. This configuration simulated a coolant channel of the MTR-type fuel element used in swimming-pool reactors. The power spectral density of the temperature fluctuation at the channel exit showed a shape similar to the power spectral density of the noise-equivalent source obtained in the Kyoto University Reactor at comparable power levels.     

Application of the Pseudo Random Binary Signals to JRR-3 High Power Dynamics Measurements

The maximum period sequence binary signal has been applied to measurements of the JRR-3 frequency response at 10 MW. Three measurements have been made to cover the frequency range of 0.003 to 10 radians/sec. The measurements were readily made when the reactor temperature was in equilibrium. Even when it was not, and the reactor power was varying slowly, a power drift compensating system made it possible to carry out accurate measurements. The results obtained under equilibrium and non-equilibrium conditions agreed well with each other.

A simple mathematical model has been used to predict the JRR-3 high power dynamics. The frequency response computed with the use of this model also agreed fairly well with the experimental data.