Neutron Noise Signals Research Articles

Discovering the causes for the change of the vibration characteristics of the core support barrel in PWR nuclear power plants: A combined investigation based on ex-core neutron noise analysis and numerical modal analysis

As a key component of pressurized water reactor (PWR) nuclear power plants, the motion of the core support barrel (CSB) has a great impact on the safety of the nuclear reactor. However, the vibration characteristics of the CSB have not yet been fully investigated and may change due to the high doses of radiation and high temperature in the reactor. In this paper, the ex-core neutron noise signals of two PWR units throughout the multiple fuel cycles were analyzed and the vibration frequency of the CSB beam mode was found, for the first time, to drop with time during the operation of PWRs. Then, the statics analysis of the CSB was conducted to preliminarily identify the potential causes. For the further clarification of such vibration frequency drop phenomenon, a numerical assembly model of PWR reactor internals, including the CSB, hold-down ring (HDR), fuel assembly hold-down spring (FAHDS), etc., was proposed with consideration of the frictional contact and fluid-structure interaction for the modal analysis of the CSB. The results showed that the decrease of the FAHDS stiffness is the main cause for the drop of the vibration frequency of the CSB beam mode in real conditions.

Operational modal analysis for characterization of mechanical and thermal–hydraulic fluctuations in simulated neutron noise

To overcome these limitations, in this work, the Frequency Domain Decomposition (FDD) of the Operational Modal Analysis (OMA) is applied to the spectral characterization of the neutron noise.In the study of the neutron noise in KWU pre-Konvoi PWRs, the Hilbert Huang Transform and the widely-used Fourier analysis are employed to extract spectral characteristics. However, both techniques are limited in decomposing the signal in the same frequency range to distinguish the contributions from different phenomena, such as the thermal–hydraulic and the mechanical perturbations. In addition, it is difficult for these two techniques to gather and present all the results in a single plot that shows the response of the core as a whole. To overcome these limitations, in this work, the Frequency Domain Decomposition (FDD) of the Operational Modal Analysis (OMA) is applied to the spectral characterization of the neutron noise. The OMA is widely used in the study of the dynamic properties of systems and structures. The FDD was performed on a series of neutron noise signals from simulated scenarios based on the transient nodal code SIMULATE-3 K. The simulations considered different single-source fluctuations and their combinations. The methodology separates in the two first singular values and singular vectors the responses due to mechanical vibrations and thermal–hydraulic fluctuation in all the frequency range, which allows distinguishing the effects of different phenomena on the spectral characteristics at the same frequency range of the neutron noise. The good performance of OMA in the present study provides promising possibilities to infer characteristics of the input excitation from the neutron noise data in PWR. Finally, the methodology shows remarkable advantages in the compilation of the results which can be utilized for monitoring purposes.

Problem Decomposition and Information Minimization for the Global, Concurrent, On-line Validation of Neutron Noise Signals and Neutron Detector Operation

This piece of research introduces a purely data-driven, directly reconfigurable, divide-and-conquer on-line monitoring (OLM) methodology for automatically selecting the minimum number of neutron detectors (NDs) – and corresponding neutron noise signals (NSs) – which are currently necessary, as well as sufficient, for inspecting the entire nuclear reactor (NR) in-core area. The proposed implementation builds upon the 3-tuple configuration, according to which three sufficiently pairwise-correlated NSs are capable of on-line (I) verifying each NS of the 3-tuple and (II) endorsing correct functioning of each corresponding ND, implemented herein via straightforward pairwise comparisons of fixed-length sliding time-windows (STWs) between the three NSs of the 3-tuple. A pressurized water NR (PWR) model – developed for H2020 CORTEX – is used for deriving the optimal ND/NS configuration, where (i) the evident partitioning of the 36 NDs/NSs into six clusters of six NDs/NSs each, and (ii) the high cross-correlations (CCs) within every 3-tuple of NSs, endorse the use of a constant pair comprising the two most highly CC-ed NSs per cluster as the first two members of the 3-tuple, with the third member being each remaining NS of the cluster, in turn, thereby computationally streamlining OLM without compromising the identification of either deviating NSs or malfunctioning NDs. Tests on the in-core dataset of the PWR model demonstrate the potential of the proposed methodology in terms of suitability for, efficiency at, as well as robustness in ND/NS selection, further establishing the “directly reconfigurable” property of the proposed approach at every point in time while using one-third only of the original NDs/NSs.

Simulation of In-Core Neutron Noise Measurements for Axial Void Profile Reconstruction in Boiling Water Reactors

This paper reports on the development and application of a method of emulating bubbly flow by generating bubbles with random sampling methods. The purpose of the modeling is that by using the simulated random two phase flow as input, one can generate “synthetic” neutron noise signals by convoluting the input with a simplified neuronic transfer function, on which the possibility of reconstructing the axial void profile from in-core neutron noise measurements can be studied by standard spectral noise analysis methods. The long term goal of this work is to elaborate methods of neutron noise analysis, by which the local void fraction in a boiling water reactor can be determined by measurements. In this preliminary stage, two methods for the reconstruction of the axial void and the velocity profiles are discussed. The first method is based on the break frequency of the neutron auto-power spectrum, whereas the second method only utilizes the information in the transit time of the void fluctuations between axial pairs of neutron detectors. A clear and monotonic relationship between the chosen observables and the two-phase flow properties was found, but an accurate determination of the void fraction requires further development and testing of the various unfolding alternatives.

Effective use of local and global component of neutron noise signals

India is pursuing the design of Advanced Heavy Water Reactor (AHWR) to achieve the objective of thorium utilization with enhanced safety and economic competitiveness. AHWR is 920MWth pressure tube type boiling water reactors with light water as coolant and heavy water as moderator. The reactor is designed to remove the core heat by two-phase natural circulation mode not only during startup, power rising or shut down cooling but also during normal full power operation. In addition, the design incorporates many passive safety features such as advanced accumulators, gravity driven cooling system for ECC and LOCA incidences respectively and passive containment cooling system for containment cooling. In many ways AHWR will be addressing stringent safety needs of the next generation nuclear reactors.Thermal hydraulics of two-phase natural circulation reactors pose stiff challenges while establishing stable operation at all power level, establishing thermal margin, thermal hydraulics related to refueling, channel burn out caused by small break LOCA, flow instability etc. In core detectors such as SPNDs are very sensitive to perturbation to neutron flux. By the definition of 1970s, neutron flux is made of the global component generated by the reactivity noise that fluctuates in phase with the whole reactor, and the local component that originates from the axially propagating perturbations. Perturbation changes the neutron field in its neighborhood, called the local component. The paper deals about effective use of SPND signal for detecting in core thermal hydraulics during steady state operation and during reactor shut down.

Decay ratio estimation in BWRs using the exponential time-scale representation

A novel method based on time-scale representations (TSR) is proposed to determine the time evolution of the linear stability parameters of a boiling water reactor (BWR) using neutronic noise signals. TSR are obtained by the affine smoothing of the Wigner time frequency representation (TFR) via an exponential kernel, the resulting TSR is called “exponential TSR” (ETSR). From the ETSR at a low scale, it is show how to estimate an instantaneous decay ratio (IDR) whose accumulated product is called the accumulated decay ratio (ACDR) closely follows the energy distribution given by a bilinear TFR valuated at the natural resonance frequency of the BWR. Therefore the peaks of the ACDR as a function of time indicate the periods of time where the energy buildups inside the BWR happen, i.e., those points in time where the nuclear reactor is having difficulty in dissipating energy out of its system. Therefore the energy buildups are directly correlated to reactor instability and, accordingly during those events the ACDR increases its value as expected indicating local instabilities. The methodology presented in this paper is validated using synthetic signals as well as real neutronic noise signals from the Laguna Verde event and the Ringhals stability benchmark.

Decay ratio estimation using the instantaneous autocorrelation function

In this paper a novel method based on the instantaneous autocorrelation function (IACF) is proposed to determine the time evolution of the linear stability parameters of a boiling water reactor (BWR) using neutronic noise signals. The IACF allow us to track the instantaneous signal envelope changes in time by defining an envelope instantaneous decay ratio ( EIDR) which is used as a measure of local stability. In order to account for long-term changes in BWR stability, the accumulated envelope-based decay ratio ( ACEDR) measure is introduced as the product of the local EIDRs. As it is shown in this paper, the ACEDR measure clearly reflects major long-term changes in BWR stability. Last the natural resonant frequency of the nuclear reactor is estimated from the Fourier transform of a weighted version of the IACF called generalized local autocorrelation function (GLACF). In order to validate our method, synthetic and real neutronic signals were used. The methodology was tested on the Laguna Verde Unit 1 and three events reported in the Ringhals stability benchmark.

Decay ratio estimation based on time–frequency representations

A novel method based on bilinear time–frequency representations (TFRs) is proposed to determine the time evolution of the linear stability parameters of a boiling water reactor (BWR) using neutronic noise signals. TFRs allow us to track the instantaneous frequencies contained in a signal to estimate an instantaneous decay ratio (IDR) that closely follows the signal envelope changes in time, making the IDR a measure of local stability. In order to account for long term changes in BWR stability, the ACDR measure is introduced as the accumulated product of the local IDRs. As it is shown in this paper, the ACDR measure clearly reflects major long term changes in BWR stability. Last to validate our method, synthetic and real neutronic signals were used. The methodology was tested on the Laguna Verde Unit 1, two events were reported in the Forsmark stability benchmark.

Decay Ratio Estimation of BWR Signals Based on Wavelet Ridges

A wavelet ridge application is proposed as a simple method to determine the evolution of the linear stability parameters of a boiling water reactor nuclear power plant (NPP) using neutronic noise signals. The wavelet ridges are used to track the instantaneous frequencies contained in a signal and to estimate the decay ratio (DR). The first step of the method consists of denoising the analyzed signals by a discrete wavelet transform to reduce the interference of high-frequency noise and concentrate the analysis in the band where crucial frequencies are presented. Next is computation of the wavelet ridges by a continuous wavelet transform to obtain the modulus maxima from the normalized scalogram of the signal. In general, associations with these wavelet ridges can be used to compute the instantaneous frequency contained in the signal and the DR evolution with the measurement. To study the performance of the wavelet ridge method, by computing the evolution of the linear stability parameters, both simulated and real neutronic signals were considered. The simulated signal is used to validate methodically and to study some features of the wavelet ridge method. To demonstrate the method applicability, three real neutronic signals related to instability events in the Laguna Verde NPP and Ringhals and Forsmark stability benchmarks were analyzed. The investigations show that most of the local energies of the signal are concentrated and that DR variations of the signals were observed along the measurements.

03/02239 Effect of uncompensated SPN detector cables on neutron noise signals measured in VVER-440 reactors: Kiss, S. et al. Nuclear Engineering and Design, 2003, 220, (2), 179–192

03/02239 Effect of uncompensated SPN detector cables on neutron noise signals measured in VVER-440 reactors: Kiss, S. et al. Nuclear Engineering and Design, 2003, 220, (2), 179–192

BWR stability monitoring using adaptive methods

Adaptive methods are system identification methods which are commonly used to analyze nonlinear and nonstationary signals. The use of three of these methods to study the time dependence evolution of the linear stability parameters of a Boiling Water Reactor using neutronic noise signals has been analyzed. Both simulated and real neutronic signals have been considered to perform the analysis.

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

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 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 2nd 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.

Instability localization with artificial neural networks (ANNs)

The aim of this piece of research is to investigate the potential of artificial neural networks (ANNs) for tackling the problem of instability localization. The instability is modeled by a variable strength absorber (point-source) in a two-dimensional bare reactor model with a one neutron-energy group. The proposed approach constitutes an exercise in simplicity in that: (1) an arbitrarily simplified model is employed for ANN training and validation; (2) few training and validation patterns of low complexity are utilized; (3) the ANN inputs are derived directly from the neutron noise signals, the proposed location of instability is given on-line via an uncomplicated combination of ANN outputs; (4) the ANN architecture is independent of the number of possible locations of instability. In fact, unlike previous approaches which employ hundreds of outputs (one for each fuel assembly), only two ANN outputs are employed representing the X- and Y-coordinates (location) of instability; (5) the responses of only a few detectors are employed; (6) a measure of confidence in the prediction is assigned. The results of ANN testing, which is performed on patterns from both actual and simplified models, are reported and analyzed.

Core Asymmetry Evaluation using Static Measurements and Neutron Noise Analysis

A case study is presented describing an analysis initiated by a real core event at Paks NPP Hungary. The analysis utilised the static data provided by the VERONA core monitoring system as well as the results from a specially organised series of measurements using the neutron noise signals from in-core Rh SPN detectors. It was demonstrated that the observed temperature rise imbalance was not caused by a power anomaly but by some partial flow blockage affecting a substantial part of the core. In addition to the static measurement results the actual coolant velocity within the affected and non-affected assemblies could be evaluated by using noise technique. By virtue of continuous analysis and monitoring the unit could be operated for a substantial part of the planned cycle length.

Burg spectrum estimation of neutron noise for the monitoring of reactor internal vibration with a hypothesis test

Traditionally, non-parameter spectrum estimation has been widely used in the analysis of neutron noise obtained from ex-core detectors for reactor structure monitoring. Because a limited data results in a severe bias, the resolution is not adequate for a monitoring application. The Burg spectrum estimation of the parameter method distinctively traces the modes of the reactor's internal structures with the order of 160 tested by χ 2 (chi) distribution and F distribution. Fuel 1st mode, 2nd mode, beam mode and shell mode in association with a reactor's internal vibration are more discernable than the spectrum estimator method proposed in ASME-OM/SG part 5 with the neutron noise signals of Ulchin 1. The result gives us the early information of the joint condition between thermal shield and core support barrel.

96/05003 Potential exposure in nuclear safety

In the frame of extensive work for the development of incipient failure detection methods, neutron noise signals in PWRs of different designs have been investigated using correlation techniques and interpreted by means of theoretical and experimental analysis of the vibrational behaviour of reactor internals. As driving forces of the dominant neutron noise sources have been identified: core barrel beam mode vibrations, shell mode vibrations of cylindrical components and fuel assembly beam mode vibrations. Variations in the long term behaviour of the ex-vessel neutron noise spectra are shown and related to changes in the mechanical properties of fuel assemblies.

Determination of Mass Flow Rates from Measured In-Core Two-Phase Flow Transit Times in a Boiling Water Reactor

By analysing the neutron noise signals of the in-core detectors, the two-phase flow transit times have been measured for unit 1 of Tarapur Atomic Power Station (General Electric Boiling Water Reactor). Our study has shown that it is reasonable to identify the fluid velocity derived from the measured transit time with the volumetric flux. This interpretation offers the advantage of determining the mass flow rate by using the measured transit time and the steam quality obtained by the heat balance. Further, it is feasible to infer gross flow rate changes by combining the calculated value of the transit time as a function of flow with the periodic measurements of transit time over the operating cycle of the reactor.

Laboratory experiments with impacting fuel rods

Vibration surveillance and diagnostics of fuel rods and fuel assemblies are important tasks in NPPs. They are so much the more difficult since the commonly used accelerometers do not operate properly under the extreme conditions of the core of reactor. Therefore vibration (or impact) of fuel rods can only be detected indirectly from neutron noise or pressure fluctuation signals. Thus accurate knowledge of vibration phenomena and measurability is very important. Experimental results on models without limiter give good coincidence with theoretical calculations. Spectra measured on impacting rod become smoother with increasing impacting level. Spectra of fuel rods have a wider range in impacting rate and higher level of smoothing than spectra of model rod have. The impacting rate strongly depends on mechanical properties of the rod. By the experiments, one can state that as for Fourier spectra the only thing caused by the impacts is the smoothening. This, however, has the very remarkable consequence, that disappearance of certain peaks does not always reflect a decrease in the vibration amplitude, but can also mean the opposite case — a vibration increased to such an extent that the rod is impacting on its surroundings. Thus there is a higher chance to give faulty diagnosis by Fourier spectra only. Consequently investigation of fuel rod vibration requires increased caution.

Determination of Mass Flow Rates from Measured In-Core Two-Phase Flow Transit Times in a Boiling Water Reactor.

By analysing the neutron noise signals of the in-core detectors, the two-phase flow transit times have been measured for unit 1 of Tarapur Atomic Power Station (General Electric Boiling Water Reactor). Our study has shown that it is reasonable to identify the fluid velocity derived from the measured transit time with the volumetric flux. This interpretation offers the advantage of determining the mass flow rate by using the measured transit time and the steam quality obtained by the heat balance. Further, it is feasible to infer gross flow rate changes by combining the calculated value of the transit time as a function of flow with the periodic measurements of transit time over the operating cycle of the reactor.