Detection of Flow-Induced Vibration of Reactor Internals by Neutron Noise Analysis

1670 IEEE TRANSACTIONS ON NUCLEAR SCIENCE, VOL. 55, NO. 3, JUNE 2008

Detection of Flow-Induced Vibrationof Reactor Internals by Neutron

Noise AnalysisSaleem A. Ansari, Muhammad Haroon, Zakaullah Sheikh, and Zafar Kazmi

Abstract—Flow-induced vibrations of reactor core componentshave been a major cause of failure of reactor internals in many nu-clear power plants. To address this issue international standards(viz., ASME-OM-05) require monitoring of structural integrity ofreactor core in nuclear power plants. A reactor internals vibra-tion monitoring system (IVMS) has been developed for nuclearpower plant surveillance. The system detects the core barrel mo-tion and flow-induced vibrations of reactor internals by analyzingthe inherent fluctuations (reactor noise) present in the neutron fluxsignals from ex-core neutron detectors. Before application of theIVMS in nuclear power plants the system hardware and method-ology/software has been extensively tested on the research reactorthrough a series of reactor noise measurements. In the noise mea-surements the neutron flux signals were correlated with the signalsfrom vibration sensors mounted on reactor structure and controlrods drive mechanisms. The frequency spectra of reactor powersignals obtained with IVMS had shown small oscillations in theneutron flux signals at well-defined frequencies. There was a strongcorrelation of these fluctuations with the establishment of coolantflow through the core. With the help of elaborate experiments usingthe IVMS the cause of oscillations in neutron flux signals was at-tributed to the flow-induced vibrations of control rods, and the par-ticular vibrating control rod was identified. The magnitude of thedisplacement of vibrating control rod was calculated from the mea-sured power spectral density of neutron power signals. It has beenshown that ex-core neutron noise measurements are more sensitivein determining the dynamic behavior of reactor internals than thevibration sensors mounted at remote, out-of-the core locations.

Index Terms—Control rod, data acquisition, flow induced vibra-tion, neutron noise, temperature fluctuations.

I. INTRODUCTION

THE fluctuation about a mean value of signals from sensorsmonitoring the status of the reactor core is called reactor

noise. Reactor noise analysis is a well established technique fordetection of malfunctions in core components by determiningtheir dynamic behavior [1]. The principle of reactor noise anal-ysis is that any mechanical or thermal-hydraulic disturbances inthe reactor core are transformed in the fluctuations in reactivityand neutron flux due to the reactivity-power transfer function.The analyses of neutron flux and other process signals thereforegive an insight into the phenomenon occurring in the core. Themajor advantage of this technique is that it does not require anyperturbation to the reactor core, since all signals are acquired atnormal, steady-state power operation. Also, in many cases no

Manuscript received January 2, 2008; revised February 15, 2008.The authors are with the Directorate of Nuclear Power Engineering-Reactor,

Islamabad, Pakistan.Digital Object Identifier 10.1109/TNS.2008.921490

additional sensors are required for noise measurements and theneutron flux, temperature, flow and pressure signals from stan-dard plant instrumentation are used. The analysis of the fluctu-ations in neutron flux signals is particularly important to get aninsight into the dynamics of in-core phenomena.

The analysis of reactor noise requires the knowledge of var-ious noise sources in the reactor that affect the reactivity of thesystem. The mechanical vibration of the fuel elements or controlrods has been identified as one major source of neutron noise,and noise analysis can be used to detect such phenomenon [2].The fluctuations in coolant temperature, flow and pressure alsocontribute to the reactivity and neutron noise in nuclear powerplants.

The reactor noise techniques have been employed in nuclearpower plants for the development of advance reactor coresurveillance systems. In pressurized water reactors (PWRs)the analysis of fluctuations in neutron flux signals from theex-core neutron detectors is used for the detection of flow-in-duced vibrations of core barrel, to meet the requirements ofASME operation and maintenance standard (ASME-OM-5).In this technique the cross power spectral density (CPSD) ofthe neutron flux signals obtained from the pairs of ex-coreneutron detectors located on diametrically opposite side of thecore is used to determine the frequencies and displacementof the core barrel motion [3]. The CPSD plots display a highcoherence and 180 phase at the frequencies of the core barrelmotion, and the magnitude of the displacement or swing ofcore barrel is obtained directly from the CPSD magnitude atthe corresponding frequencies.

II. REACTOR NOISE MEASUREMENTS

A Reactor Internals Vibration Monitoring System (IVMS)has been developed for reactor core surveillance in nuclearpower plants, based on the methodology described in Section I,[4]. The hardware, software and detection methodology ofIVMS have been tested by installing the system at the Pak-istan Research Reactor (PARR-1) and making extensive noisemeasurements on the reactor. Research reactors provide anideal tool for testing system performance under a wide range ofoperating conditions (viz., low- and full-power operation, oper-ation without flow at low-power, and non-standard sequencesof control rods movement in the core), which are not possiblein operating power plants. This aspect would be described indetail below.

PARR-1 is a 10-MW pool-type reactor with MTR (MaterialsTest Reactor) fuel elements, five shim control rods and graphite

0018-9499/$25.00 © 2008 IEEE

ANSARI et al.: DETECTION OF FLOW-INDUCED VIBRATION OF REACTOR INTERNALS BY NEUTRON NOISE ANALYSIS 1671

and beryllium reflector. Earlier, the reactor had a maximumpower level of 5 MW, and the core was fuelled with high-en-riched-uranium (HEU) fuel elements. Preliminary reactor noisemeasurements were made in the past [5] on the old 5-MW re-actor core. Later, the reactor core was converted to Low-En-riched-Uranium (LEU) fuel and its power up-graded to 10 MW.This paper describes the results of neutron noise measurementsmade in the upgraded reactor using the IVMS to determine theinteraction of coolant flow with core components and its effecton reactivity and power variations in the core.

A. Interfacing of Neutron Noise and Vibration Signals WithIVMS

For the reactor noise experiments at PARR-1 a total of sixneutron flux & mechanical vibration signals were interfacedwith the IVMS. At PARR-1, the reactor power is measured withthe help of five ex-core neutron ionization chambers. Two de-tectors provide signals to two linear-flux monitoring channels,which are a part of the automatic power control system. Threeex-core neutron ion chambers provide signals to the three safetychannels of the reactor protection system. For the neutron noisemeasurements the signals from the two non-safety linear-fluxmonitoring channels (Lin-A and Lin-B), available at the reactorsignals output panel in the main control room, were connectedwith the IVMS. In addition to the neutron flux signals, four vi-bration accelerometers were mounted on the core support towerand the control rod drive mechanisms to monitor the vibrationbehavior of reactor internals. These vibration signals were alsoconnected with the IVMS. In one experiment the vibration ac-celerometers were also mounted on the drive mechanisms of theneutron detectors to correlate the flow-induced vibrations of theinstrumentation tubes with the neutron noise signals.

B. Reactor Operation

All noise measurements made with the IVMS at PARR-1were at steady state power level at low-power (below 100 kW)and full-power (10 MW) modes. It was ascertained in initialmeasurements that the pattern of neutronic fluctuations at highpower operation did not show any significant variations fromthat at lower power, and the normalized root mean square valueof power noise was almost independent of power level. Inorder to avoid any possible reactor safety implications (espe-cially for the measurements involving mounting of vibrationsensors), most of the measurements and analyses were madeat low-power. The effect of interaction of coolant flow withreactor core components was determined by acquiring neutronnoise and vibration signals with and without the forced coolantflow through the core. At PARR-1 it is possible to operatethe reactor without forced flow at low-power (below 100 kWpower), and the reactor core is cooled with natural conventionalcooling. This is a very significant application of research reac-tors for development of reactor noise techniques, since such anoption is not available in power reactors. Another feature of theresearch reactor was utilized in some experiments by makingthe reactor critical in different control rods configurations, viz.,by either inserting fully or withdrawing one control rod at a timeand operating the reactor with the remaining four control rods.Typical reactor operation time for one set of measurements was

Fig. 1. Functional diagram of signal acquisition and processing modules ofIVMS.

two-hours. The data taking at steady state power operation overthe two-hour time period enabled neutron noise spectra to beacquired with high statistical accuracy, and any local variationsin reactor power were averaged out. The reactor was operatedunder manual control to avoid the additional feedback of thecontrol system. Various reactor operation conditions employedin neutron noise experiments are shown below.

Reactor Power Modes:

High-power 10 MW.

Low-power 100 W-90 kW.

Coolant Flow Modes:

Full-flow 960 m /h.

Flow-off Natural convectional flow.

Control rods configuration:

Normal mode All control rods (shim rods 1-5)leveled at identical withdrawalpositions.

Experimental mode One control rod fully inserted or fullywithdrawn. Reactor made critical withremaining four rods.

Reactor control Manual Control Mode.

C. IVMS Signal Acquisition and Processing Modules

The neutron noise and vibration signals at PARR-1 were ac-quired and processed with the help of high-precision signal con-ditioning and data acquisition instrumentation modules of theIVMS. The block diagram showing the sensors signals and var-ious hardware modules of the IVMS used in the noise measure-ments are shown in Fig. 1. The hardware comprised of the fol-lowing modules.

• Signal Protection Module (SPM): This module com-prised of isolation amplifiers to provide protection to


TABLE ITYPICAL DATA OF NOISE SIGNALS AND INSTRUMENTATION

SETTINGS USED IN THE NOISE MEASUREMENTS

reactor signals against short-circuiting or over-voltageduring noise measurements.

• Signal Conditioning Module (SCM): This modulecomprised of instrumentation for ac-coupling, low-passfiltering and band-pass amplification of noise signals.

• Signal Acquisition Module: Comprising of 16-bit PCplug-in ADC with programmable gain for simultaneousacquisition of multi-signals.

• Data Processing Computer: The computer contained ad-vanced algorithms for signal analysis, which included: FastFourier transform (FFT); auto and cross power spectraldensity; correlation and coherence functions; digital fil-tering; frequency response function; phase & magnitudeplots.

III. RESULTS

A. Frequency Composition of Reactor Power Signals

The power spectral density (PSD) plots obtained with the helpof IVMS of neutron flux signals from Lin-A and Lin-B fluxchannels at 10 kW power are shown in Fig. 2. The figure showsfrequency contents of the neutron noise in the low-frequency(0.1–10 Hz) range obtained with a high resolution. The PSDis plotted in units of [Watt] , obtained by dividing the fluctu-ating voltage signals of linear flux channels with the signal dcvalues and normalizing the ratio to the operating power level.The coolant flow rate through the core was maintained at thenormal flow value, and all control rods were in their normal crit-ical position in the core. It is evident from Fig. 2 that the PSD ofthe reactor power fluctuations has a definite line (or resonance)structure. About 46% of the total signal power of the neutronflux signals in the 0.1–10 Hz is contained in two prominent peakregions; the first resonance having a frequency of 0.7 Hz, whilethe other peak occurring at 1.48 Hz frequency is the second har-monic of the first peak. Repeated measurements of neutron fluxsignals made in the reactor at different times under the sameoperating conditions gave identical results as of Fig. 2, thus es-tablishing the stationary nature of neutron noise spectrum. Theneutron noise pattern of the two ex-core neutron flux detectors,

Fig. 2. Power spectral density plots of neutron power fluctuations at 10 KWpower obtained from the linear flux channels signals (Lin-A and Lin-B).

Fig. 3. Coherence function between Lin-A and Lin-B signals at 10 KW power.

Lin-A and Lin-B, showed nearly identical features. The coher-ence function of Lin-A and Lin-B signals is plotted in Fig. 3.The two signals show good, near unity, coherence at low fre-quencies. The high coherence in the 0.1–1 Hz frequency rangemay be attributed due to the correlated temperature fluctuationscausing common-mode reactivity and neutron flux fluctuationsin the two detectors signals. In the 1–2 Hz bandwidth the highcoherence between Lin-A and Lin-B signals may be the resultof other reactivity noise sources, such as the flow-induced vibra-tions of reactor structure. A high coherence of the two neutrondetectors’ signals clearly indicates that the fluctuations arisingin the detectors signals are due to some global phenomena oc-curring in the reactor core, with minimal contribution from theun-correlated noise in the individual detectors (detector noise).Fig. 4 shows the phase angle between Lin-A and Lin-B signalsas a function of frequency. It is evident from the figure that thephase angle varies around zero degrees in the 0.1–10 Hz range.The reason for zero phase difference between Lin-A and Lin-B


Fig. 4. Phase angle versus frequency between Lin-A and Lin-B signals.

Fig. 5. Vibration spectrum of accelerometer signal mounted on the upper sec-tion of control rod-3(SR-3).

signals is that since the two detectors are located very close toeach other the global variations in neutron flux in the core willarrive almost simultaneously at the two detectors, due to pointreactor kinetics.

The results of several neutron noise measurements made atdifferent power levels in the 0.1–90 kW range and at 10 MWpower showed that the resonances in the neutron noise spectrumpersisted at all power as long as the coolant flow was maintainedthrough the core. The power spectral density of neutron fluxsignals varied roughly as the square of the reactor power andthe rms magnitude of neutron noise varied proportionally withreactor power. This is a well known behavior of neutron noise[6], and it allows the reactor noise to be fully investigated atlow-power levels without undue risk of reactor safety envisagedat full power.

The line or resonance structure of the PSD of neutron noisepoints to a strong oscillatory behavior of reactivity and power.The rms magnitude of the oscillations in reactor power at10 MW in 0.1–10 Hz bandwidths was calculated to be 34 kW,or 0.34% of the steady-state reactor power.

The frequency spectra of vibration signals from the four ac-celerometers mounted at various locations on reactor structurewere also analyzed. Fig. 5 shows a typical vibration frequencyspectrum obtained from the accelerometer placed at the con-trol rod drive mechanism (CRDM) of control rod -3(SR-3). Itis evident from the figure that the vibration spectra from theaccelerometer did not show any of the frequencies of the re-actor power oscillations. A similar pattern of vibrations of re-maining four control rods was also observed. One reason for themismatch in the frequencies of the vibration and neutron noisespectra may be attributed to the complex structure of CRDM of

Fig. 6. Control rod drive system components.

Fig. 7. Frequency spectra of reactor bridge vibrations.

the reactor. This is evident from Fig. 6, which shows the detailsof the CRDM components. Since it was not possible to mountthe accelerometers on the control blade (which traveled insidethe reactor core), these sensors were mounted on the outer tubesin the upper section of the CRDM at locations above the waterlevel of the reactor pool. It is therefore possible that the frequen-cies of the vibrating control blades in the core may not be trans-mitted to the sensors on the upper structure of the CRDM. Sincethe reactivity variations in the core are caused by movement (vi-bration) of control blades, hence the frequencies of reactivityand power fluctuations may not be the same as the frequenciesof mechanical vibration signals of Fig. 5. Fig. 7 shows the fre-quency spectrum of vibration of the reactor structure under fullflow, obtained from the accelerometer mounted on the core sup-port tower. Here again the vibration spectrum is markedly dif-ferent from the neutron flux spectrum.


Fig. 8. Comparison of frequency spectra of reactor power fluctuations at 1 KWwith and without forced flow.

In order to fully understand the causes of reactivity and poweroscillations in the reactor a number of elaborate experimentswere devised. The results of these experiments are described inthe following sections.

B. Effect of Coolant Flow on Neutronic Fluctuations

A comparison of the relative power spectral densities ofneutron noise obtained at 10 kW power level with no forcedflow (natural convectional cooling) and with full flow of 960m /h through the reactor core is shown in Fig. 8. Reactor powerand other operating conditions for these measurements werethe same as in the measurements shown in Fig. 2. It is evidentfrom Fig. 8, that the establishment of coolant flow in the corehad a drastic effect on neutron noise spectrum. As the flow wasswitched off the oscillations in reactor power disappeared andthe PSD of neutron noise assumed the shape of a bandwidthlimited white noise function. The rms magnitude of neutronicfluctuations without flow dropped to almost 1/15th of the rmsvalue at the same power with full flow.

In order to rule out the possibility of the flow-induced vibrationof the detector support structure of the ex-core neutron ionizationchambers to be the cause of oscillations in the detectors currentsignals (due to micro phonicsgeneration),oneaccelerometer wasmounted on the detector support assembly near the detector lo-cation. It was observed that the mechanical vibration of detectorsupport structure was negligibly small (close to the minimumdetection range of accelerometer sensitivity) without any peakstructure. These measurements, coupled with high coherence inneutron detectors signals ruled out the possibility of generationof local micro phonics in the detectors signals.

The strong influence of coolant flow on the frequencies andmagnitude of neutron power oscillations points to the flow-in-duced reactivity disturbances in the core. As mentioned in Sec-tion I the two possible sources of flow-related reactor noise are

the coolant temperature fluctuations in the core and the me-chanical vibration of the fuel elements or control rods. The fol-lowing sections describe the mathematical modeling of the noisesources made in an attempt to explain the oscillatory phenom-enon in reactor power.

C. Reactivity Effect of Coolant Temperature Fluctuations

In order to explain the fluctuations in the coolant temperaturein the upgraded PARR-1 core the one dimensional thermal-hy-draulic model of reactor core, based on the formalism ofRobinson and Van Uitert and Van Dam [7] was used. Thismodel postulates that as the coolant travels through the reactorcore the fluctuations in inlet coolant temperature are trans-formed into the fluctuations in the core average temperature.These core average temperature fluctuations, in turn, producereactivity and power fluctuations through respective coolantand fuel temperature coefficients of reactivity. The fluctuationsin inlet coolant temperature are assumed to be random in nature(white-noise). In PARR-1, the temperature feedback affect maybe neglected at low power. The PSDs of reactivity and neutronnoise in the core due to inlet temperature fluctuations are givenby the relations [7]

(1)

(2)

where

PSD of core inlet temperature;

transfer function relating the fluctuation in theinlet coolant temperature to the core-averagedcoolant temperature fluctuations;coolant temperature coefficients of reactivityrelating the average coolant temperaturefluctuations to the reactivity fluctuations;zero-power transfer function relating the reactivityvariations to the variations in neutron density, orreactor power.

The effect of fuel temperature fluctuations on neutron noise isvery small and is neglected. The transfer function A/C is givenby the relation

(3)

where

flux gradient;

reflector saving;

height of core;

function in Laplace domain given by

(4)

being the coolant velocity through the core, and are thecharacteristic break frequencies of coolant and fuel temperaturefluctuation, respectively.


Various heat transfer and neutronic parameters of PARR-1core required in (1) to (4) for the calculation of neutron noise dueto coolant temperature fluctuations were taken from the reactordesign report [8].

The calculated PSD of neutron noise caused by the white-noise fluctuation in inlet coolant temperature is shown in Fig. 9.The behavior of the reactor noise is that of bandwidth limitedwhite noise in the 0.1–1 Hz frequency range. The sink structureof the temperature frequency spectrum as predicted by Kosaly[9] is visible. It is therefore concluded that the coolant temper-ature fluctuations in the core do not result in any oscillations inreactor power.

D. Effect of Control Rods Vibrations on Reactivity Noise

As explained in above section, the analysis of temperaturefluctuations could not explain the resonance structure in the neu-tron noise spectra that appeared with the establishment of thecoolant flow through the core. The effect of the other probablenoise source on reactor power oscillations, viz., control rods vi-brations, was investigated through a series of elaborate experi-ments. The objectives of these experiments were two-fold: (a)To eliminate possible vibrations in one individual control rodby cutting off the coolant flow through that particular rod, whileother rods were subjected to full flow; (b) To minimize the re-activity effect of one control rod at a time so that the vibrationsin this rod did not cause perturbation in reactor power fluctua-tions. The first objective was achieved by making use of the de-sign of the control rod drive mechanism of PARR-1. Accordingto the design, the control rod blade moved in an annulus insidethe control fuel element (CFE). When a particular rod was fullyinserted in the CFE, the shock absorber structure at the top ofthe control blade blocked the flow of coolant through the an-nulus and there was no flow around the control blade. This op-tion of full rod insertion is only allowed at low-power operationto avoid heating effect due to non-flow in control blade annulus.In this way it was possible to eliminate the flow-induced vibra-tions in one control rod. In the noise experiments one controlrod was fully seated at a time and reactor was made critical withremaining four control rods. The reactor noise signatures wereacquired with the IVMS and the process was repeated with allfive control rods. It is worth mentioning that although in thisparticular configuration the vibrations in the individual rod areexpected to be minimized, maximum negative reactivity of therod is inserted in the core due to full rod insertion. The secondobjective described above was achieved by fully withdrawingout of the core one control rod at a time, so that it had no reac-tivity effect and hence no contribution to the power noise. Thereactor was made critical with remaining four control rods. Inboth sets of these experiments it was expected that as the impactof vibration of a particular rod was minimized the reactivity andpower oscillations should disappear, resulting in the identifica-tion of that rod. Since the experiments required special reactoroperations, the reactor power was maintained at 10 kW, to meetsafety requirements.

The effect on neutron power noise of flow blockage throughindividual control rods (full insertion) at 10 kW power level anda flowrate of 960 m /h were analyzed for all five control rods. Itwas observed that insertion of control rods 1, 2, 4, and 5 had nosignificant effect on the peaks in the neutron noise signatures,

Fig. 9. Calculated temperature PSD due to (1oc) coolant inlet temperaturefluctuations.

and the neutron noise spectra had the same shape as Fig. 2, ob-tained for all rods leveled at the same position. However, whencontrol rod-3 was fully inserted in the core the oscillations inreactor power disappeared and frequency spectrum of the re-sulting reactor noise became free of any resonances. This be-havior is evident from Fig. 10, which shows the neutron noisespectra obtained with control rod-3 and control rod-1 fully in-serted. Fig. 11 shows the frequency spectra of the neutron noiseas these two control rods were fully withdrawn out of the core(one at a time). This figure also shows the same pattern as ofFig. 10, and the withdrawal of control rod-3 resulted in the re-duction in magnitude of oscillations in reactor power.

The two sets of experiments performed at PARR-1 using theIVMS confirm that the oscillations present in reactor power aregenerated by the flow-induced vibrations in control rod -3, andother core components do not experience any significant vibra-tions due to interaction of coolant flow.

From the experimental results shown in Figs. 8–11 it is alsopossible to determine the displacement magnitude of controlrod-3 due to flow-induced vibration. This was done in the fol-lowing way.

• Fig. 8 (Curve-I) represents the magnitude of power spec-tral density of neutron noise for the case when all controlrods were withdrawn to their normal critical level and theneutron noise was composed of vibrations and other noisesources. The magnitude of PSD of Fig. 8 was computedin the resonance frequency region (0.1–1.5 Hz). From thisvalue, the PSD magnitude of Curve-II of Fig. 10 (no vi-brations) was subtracted in the same frequency region toobtain PSD of neutron noise excited solely by SR-3 vibra-tion

(5)


Fig. 10. Frequency spectra of neutron noise obtained by fully inserting SR-1and SR-3, in the core.

Equation (5) is valid since different noise sources add inquadrature.

• From , the root mean square magnitude ofneutron noise due to vibration was obtained byusing relation, . The rms value of

was transformed into rms of reactivity fluctuationsby dividing the former with the open-loop reactor transferfunction defined in Section III-B.

• From the reactivity variations thus obtained, the rms dis-placement of the control rod in the core, , required toproduce the particular reactivity effect was determined inthe following way.— The differential reactivity worth of control rod

SR-3 was measured at its normal critical position inthe core. The differential rod worth is the change in re-activity, , introduced in the core by a slight varia-tion in the rod position in vertical direction, x; i.e.,

x.— Due to similar nature of the distributions of neutron

fluxes in the radial and axial directions in PARR-1 core,and due to very small radial (horizontal) displacementof the control rod as a result of flow-induced vibration,the reactivity effect of the control rod displacement inthe vertical and radial direction was assumed to be sim-ilar. The measured value of differential rod worth in thevertical direction could therefore be used to calcu-late rod displacement in the radial direction.

The relation for the computation of displacement magnitude ofcontrol rod vibration then becomes

(6)

Fig. 11. Frequency spectra of neutron noise obtained by fully withdrawingSR-1 and SR-3, out of the core.

where

Measured differential reactivity worth (reactivityper cm of rod withdrawal) of control rod SR-3 at itscritical position for the present core.

Zero-Power Reactor Transfer Function

Using (5) and (6), the peak-peak displacement of control roddue to flow-induced vibration was computed as

mm (7)

It may be mentioned that a vibration displacement of about 1mm magnitude is still negligibly small than the clearance avail-able in the annulus around the control blade in the control fuelelement. Hence, there is no chance of impact of control bladewith the annulus wall due to the flow-induced vibrations.

The calculated neutron PSD due to temperature fluctuation(Fig. 9) was compared with the experimental noise spectrumwithout rod vibration (Fig. 10, curve-II). The shapes of the twospectra agree quite well within the frequency region 0.1 to 3.5Hz. Dividing the two spectra within this frequency range gavean estimate of the root mean square value of coolant inlet tem-perature fluctuation as

C


IV. CONCLUSION

The application of Reactor Internals Vibration MonitoringSystem developed for nuclear power plant surveillance has beendemonstrated in a research reactor for detection and identifica-tion of flow-induced vibrations of control rod. The magnitudeof the displacement of vibrating control rod has also been cal-culated from the measured power spectral density of neutronnoise. The measurements have provided experimental valida-tion of neutron noise technique for detection of flow-inducedvibrations of in-core components.

REFERENCES

[1] R. F. Saxe, “Dynamic analysis of a pressurized water reactor,” Ann.Nucl. Energy, vol. 3, pp. 477–488, 1976.

[2] D. Wach, “The analysis of at-power neutron flux noise in the frequencyrange of vibrating reactor structures,” Ann. Nucl. Energy, vol. 2, pp.353–363, 1975.

[3] D. N. Fry, J. March, and F. J. Sweeny, “Use of neutron noise for diag-nosis of in vessel anomalies in light water reactors,” ORNL/TM-8774,Oak Ridge Nat. Lab., Jan. 1984.

[4] A. Trenty, “Operational feedback on reactor internals structure vibra-tions in 54 French PWRs during 300 fuel cycles,” in Symp. NuclearReactor Surveillance and Diagnostics, SMORN-VIII, 1995.

[5] S. A. Ansari and S. K. Ayazuddin, “Detection of control rod vibrationby neutron noise analysis,” Nucl. Technol., vol. 67, pp. 9–13, 1984.

[6] J. C. Robinson, “Analysis of neutron fluctuation spectra in the OakRidge research reactor and the high flux isotope reactor,” ORNL-4149,Oak Ridge Nat. Lab., Oct. 1967.

[7] G. C. Van Uitert and H. Van Dam, Analysis of Pool Type Reactor NoiseProgress in Nuclear Energy. New York: Pergamon, 1977, vol. 1, pp.644–662.

[8] Preliminary Safety Analysis Report of PARR-1, 2001, revised.[9] G. Kosaly and M. M. R. Willams, “Point theory of neutron noise in-

duced by inlet temperature fluctuations and random mechanical vibra-tions,” Atomkernenergie (ATKE), vol. 18, no. 3, 1971.

[10] S. E. Stephoson, D. P. Roux, and D. N. Fry, “Neutron fluctuatingspectra in the oak ridge research reactor,” ORNL-TM-1401, Oak RidgeNat. Lab., May 1966.

Detection of Flow-Induced Vibration of Reactor Internals by Neutron Noise Analysis

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