NUREG/CR-5479

ORNLITM-11398

Current Applications of

Vibration Monitoring and

Neutron Noise Analysis .6

Detection and Analysis of Structural

Degradation of Reactor Vessel Internals

from Operational Aging

Prepared by B. Damiano, R. C. Kryter

Oak Ridge National Laboratory

Prepared for

U.S. Nuclear Regulatory Commission
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NUREG/CR-5479

ORNIJTM-11398

RV

Current Applications of

Vibration Monitoring and

Neutron Noise Analysis

Detection and Analysis of Structural

Degradation of Reactor Vessel Internals

from Operational Aging

Manuscript Completed: February 1990

Date Published: February 1990

Prepared by

B. Damiano, R. C. Kryter

Oak Ridge National Laboratory

Operated by Martin Marietta Energy Systems, Inc.

Oak Ridge National Laboratory

Oak Ridge, TN 37831

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN B0828

Under Contract No. DE-AC05-840R21400
ABSTRACr

Monitoring programs using vibration monitoring or neutron noise analysis have

demonstrated the ability to detect and, in some cases, diagnose the nature of reactor

vessel internals structural degradation. Detection of compromised mechanical integrity

of reactor vessel internal components in its early stages allows corrective action to be

taken before major weakening or damage occurs. In addition to the economic benefits

early detection and correction can provide, they can also help maintain plant safety.

Information on the condition of reactor vessel internal components gained from a

monitoring program supplements in-service inspection results and may be useful in

justifying plant license extension.

This report, which was prepared under the Nuclear Plant Aging Research Program

sponsored by the U.S. Nuclear Regulatory Commission, discusses the application of

vibration monitoring and neutron noise analysis for monitoring light-water reactor vessel

intemals. The report begins by describing the effects of structural integrity loss on

internals vibration and how measurable parameters can be used to detect and track the

progress of degradation. This is followed by a description and -comparison of vibration

monitoring and neutron noise analysis, two methods for monitoring the mechanical

integrity of reactor vessel internal components. The major section of the report

describes the status of reactor vessel internals condition monitoring programs in the

United States, Federal Republic of Germany, and France, three countries having

substantial commitments to nuclear power. The last section presents guidelines for U.S.

utilities wishing to establish reactor internals condition monitoring programs.

Hi.
CONTENTS

Page

ABSTRACT ........ 1..................................iii

LIST OF FIGURES ......................................... vii

LIST OF TABLES .......................................... ix

1. INTRODUCTION ....................................... 1

2. EFFECTS OF LWR INTERNALS MECHANICAL DEGRADATION

ON VIBRATION SIGNATURES ............................. 3

3. METHODS OF SURVEILLANCE ............................ 7

3.1 Similarities Between Mechanical Vibration Monitoring and

Neutron Noise Analysis ................................. 7

3.2 Advantages and Disadvantages of Mechanical Vibration

Monitoring and Neutron Noise Analysis ...................... 8

4. SUMMARY OF OPERATING EXPERIENCE .................... 10

4.1 U.S. Experience . ...................................... 10

4.2 Oerman Experience .................................... 11

4.3 French Experience .................................... 16

4.4 Cost Effectiveness ..................................... 19

5. UTILITY GUIDELINES FOR THE IMPLEMENTATION OF

INTERNALS MONITORING PROGRAMS ..................... 21

6. CONCLUSIONS ......................................... 24

ACKNOWLEDGMENT ...................................... 25

REFERENCES ......... 26

v
LIST OF FIGURES

Figure Page

1. PWIR and BWR internals .................................... 5

2. Vibration and noise sensor positions of a four-loop German PWR ...... 12

3. Modeling of PWR primary circuit and reactor internals .... .......... 14

4. Monitoring system in French PWRs equipped with partial

thermal shield ........... ................................ 17

5. SAFRAN test loop ....................................... 18

vii
LIST OF TABLES

Table Page

1. Examples of LWR components experiencing failure or excessive

flow-induced vibration during commercial operation ................. 4

2. Methods that have been used to detect failure or degradation

of specific reactor internal components ......................... 8

3. Advantages and disadvantages of reactor noise program

organization ............. ............................... 22

4. Desired knowledge for reactor noise analysts ..................... 23

ix
1: INTRODUCION''

The detection of light-water reactor (LWR) internals structural degradation caused

by operational aging may become more important to U.S. utilities as the median age of

U.S. nuclear power plants increases. It has been demonstrated that monitoring programs

using vibration monitoring or neutron noise analysis can effectively detect the loss of

mechanical integrity of reactor internal components at an early stage before severe

damage occurs. Such programs have the potential to reduce plant downtime and its

associated costs, to make periodic maintenance more effective, and to maintain a high

level of plant safety. Another potential use for the information gained from an internals

condition monitoring program may be as an aid in justifying extension of a plant's

operating license.

Monitoring of reactor internals structural integrity can be considered an

application of the general concept of predictive maintenance. Predictive maintenance is

widely used by industry to monitor the condition of rotating machinery, where it has

proved to be cost effective.' It is assumed that the'application of predictive

maintenance techniques for monitoring the condition of reactor internals would similarly

become widespread if comparable benefits could be demonstrated. Although some

internals monitoring programs have shown mixed results, others have proved

unquestionably beneficial and cost effective.

It has been stated that internals monitoring programs in Europe are five to ten

years ahead of those in the United States.2 If viewed as an opportunity rather than a

criticism, this situation allows U.S. utilities to benefit from the considerable European

experience base, where vessel internals condition monitoring has been integrated into

regular plant maintenance programs in many cases. Thus, it should be possible for U.S.

utilities to implement effective monitoring programs with a minimum of experimentation.

The information presented in this report was obtained from several sources. The

main source is the series of Specialists' Meetings on Reactor Noise which have been

held every two to three years since 1974. Other sources include papers from other

meetings and conferences, information from contacts within the nuclear industry, and

information gained by Oak Ridge National Laboratory (ORNL) staff involved in

research on the application of neutron noise analysis for monitoring the continued

mechanical integrity of reactor internal components. The authors recognize that there

may be significant work of which we are unaware. However, the information on which

this report is based is believed sufficiently complete to describe accurately current

applications of both vibration monitoring and neutron noise analysis for monitoring the

condition of reactor internals.

This report, which was prepared under the Nuclear Plant Aging Research Program

sponsored by the U.S. Nuclear Regulatory Commission (USNRC), discusses the

application of vibration monitoring and neutron noise analysis for monitoring LWR

vessel internals. The report is organized as follows. First, the effects of internals

mechanical degradation on the nature of sensible vibration signatures are reviewed. This

is followed by a brief description of vibration monitoring and neutron noise analysis,

which includes a comparison and evaluation of these two methods. Next, current

practices are summarized, and example applications of these methods are given for both

1
2

the United States and Europe. The report concludes with a summary of guidelines for

establishing what the authors consider a reasonable internals monitoring program for

U.S. utilities.
2! EFFECTS OF LWR INTRNAIS MECHANICAL DEGRADATION ON

VIBRATION SIGNATURES

Flow-induced vibration of reactor internal structures has been and continues to be

a problem in commercial LWRs. A partial list of internal component abnormalities or

failures attributable to excessive vibration is given in Table 1. In this report, internals

degradation means structural integrity deterioration that does not necessarily prevent a

structure from performing its intended function. Using this definition, examples of

degradation include weld cracking or loss of bolt preload. Failure is reached when and

if the degradation progresses to the point where the structure "breaks" in such a way

that it can no longer perform its function.

Many cases of internals degradation are caused by excessive flow-induced vibration.

As coolant is forced through the core, internal structures are-subjected to strong

fluctuating pressure and drag forces (see Fig. 1). These forces inevitably excite some of

the natural vibratory modes of the internal structures, resulting in motion that may

eventually degrade the structural integrity of the affected components. A component

does not necessarily require direct hydraulic excitation to experience vibration. Many

components are mechanically coupled; thus, excitation of one component may result in

vibration of a significant portion of the internals.

Often a mechanical degradation will affect the resonant frequencies of internal

structures; thus, a resonant frequency shift may be used to detect some forms of

mechanical degradation. Two forms of mechanical degradation that often result in

resonant frequency shifts, loss of bolt preload, and fatigue are briefly described in the

following paragraphs.

Loss of bolt preload may result in abnormally large vibration amplitudes and

shifted resonant frequencies and often precedes more severe structural damage. As a

connection loosens, the stiffness and mechanical coupling between connected

components is reduced; this often changes both the frequency and amplitude of the

vibration because of changes in the stiffness and effective mass of the structure.

Greater vibration amplitude results in greater stresses within the internal components,

thereby accelerating the progression of damage. In structures joined by multiple

connections, loss of preload in one connection increases the load carried by those

remaining, possibly increasing the rate at which other connections loosen and fatigue.

The accompanying frequency changes can be used to detect and ascertain the nature of

these degradations.

Fatigue damage can occur when a material is subjected to alternating stresses.

The rate at which fatigue damage progresses is affected primarily by the mean stress

level, the frequency of stress reversal, the magnitude of the alternating stress, and the

material's fatigue strength. As damage progresses, fatigue cracks grow, reducing the

load-carrying portion of the structure. As in loss of preload, the remaining intact

portion then carries greater mean and alternating stresses, resulting in an accelerated

rate of fatigue-crack growth. As fatigue cracks grow, the connecting stiffness decreases.

The progression of such fatigue damage can be detected from the resulting frequency

shift.

3
4

Table 1. Eiampis of LWR n hponents cpeincing failure .r or e

rwmnduced bratoin during cxnmercial opeacion

Reactor Component Problem

KKS Stade Core support Loose bolts in structure

below core barrel

Palisades Core barrel flow-induced core barrel

motion

Novovoronezh-1 Thermal shield Loosening of thermal

KWO Obrigheim shield

Oconee 1,2,3

St. Lucie

Millstone

Maine Yankee

Trino Vercelles

Trojan Fuel elements *Baffle-jetting"-damaged

fuel elements on core

periphery

Crystal River Control rod spider Broken spider caused by

flow-induced vibration of

control rods

BWR4s Instrument tubes Damaged fuel channel

boxes caused by impacting

with instrument tubes

Millstone-I Feedwater sparser Vibration and cracking

observed

Pathfinder Steam separators Vibration-induced

structural failure
5

Control rod

assembly

Plenum

assembly

Upper grid

Outlet nozzle

Core barrel

Lower grid

Flow distributor

Plenum cover

Control rod

guide tube

Core support

shield

Inlet nozde

Fuel assembly

Core support

assembly

'Reactor vessel

Thermal shield

I Guide lug

-Incore

instrument

tubes

(a) PWR internals

Fi. 1. PWR and BWR internals (Source: J. A. Thie, Power Reactor Noise, American Nuclear

Society, La Grange Park, I1L, 1981. Reprinted by permission.)
6

Reactor vessel

Steam outlet nozzle

Vessel flange

-St Seam pressure

measurement tap

Steam separator

- Feedwater inlet nozzle

Water level

measurement taps

Core spray sparger -

Upper core grid -

In-core flux monitor

assembly

Fuel assembly

- Control rod

, Jet pump body

Lower core grid

Recirculating water

Control rod guide tube- outlet nozzle

Core differential

pressure tap

-Vessel support skirt

In-core

sleeve Control rod drive

housing

(b) BWR internak

Fi. 1. (continued)
3. METHODS OF SURV3ELLANCE

Although the presence of structural degradation may result in both increased

vibration amplitude and a shift in vibration frequency, the latter is the primary means for

detecting degradation. Measurement of reactor internals vibration most often is

performed using mechanical sensors mounted on the reactor vessel or other primary

system components, or by analyzing the noise content of the signals generated by excore or in-core neutron flux detectors. Table 2 lists the major surveillance methods that

have been used to detect the degradation of specific components.

3.1 SIMILARITEES BETWEEN MECHANICAL VIBRATION MONITORING AND

NEUTRON NOISE ANALYSIS

The form of data collected using mechanical vibration monitoring or neutron noise

analysis is similar, and data analysis techniques for both types of sensors are basically the

same. Unlike in-service visual inspections, both methods are nonintrusive; monitoring

requires no process perturbations or special conditions and in fact is best performed at

full power during normal plant operation. Although the data may be analyzed in either

the time or frequency domain, for purposes of monitoring reactor internals vibration,

analysis in the frequency domain is preferable since the frequency shifts indicative of

vessel internals degradation are easily detected.

Frequency domain analysis typically involves using a fast Fourier transform (FFT)

algorithm to convert signals from the time to the frequency domain. The results may be

displayed as an auto-power spectral density (APSD) for each signal or, if multiple signals

are sampled, -is cross-power spectral densities (CPSDs) and coherences for each signal

pair. Examination and comparison of resonant peak amplitudes and frequencies of

APSDs and CPSDs and values of CPSD coherence and phase have proved useful in

detecting and diagnosing anomalies.

Since the detection process involves comparison of spectral features, a baseline or

reference signature representing normal operation is an essential starting point. An

effective monitoring program, therefore, entails regular data collection, comparison, and

analysis to ensure that anomalies will be detected before major damage occurs. Baseline

signatures may need to be updated periodically, for example, if modifications to the

vessel internals are introduced.

It should be noted that baseline signatures may not remain constant over an entire

fuel cycle even though the reactor is unaltered mechanically and is operating normally.

For example, neutron noise signatures are often affected by fuel burnup and control rod

position, both of which vary continuously during a fuel cycle. Similarly, vibration

signatures may be influenced by reactor temperature, pressure, water level, and power

level. Thus, the dependence of the baseline noise signatures on parameters other than

the mechanical vibration characteristics of the internals must be well understood. These

considerations indicate the need for a long-term monitoring program which, ideally,

should continue throughout the life of the plant, beginning during construction and

initial testing.

7
8

Table 2. Methods that have been used to detect failure or

degradation of specific reactnr internal cxnpets 3

Surveillance Method

Neutron

Vibration noise Visual

Component monitoring analysis inspections

Thermal shield X X X

Core support X X

Core barrel X X X

Fuel assemblies X X X

Control rods X

Instrument thimbles X X

Indicates method used for degradation or failure detection.

tVisual inspection detected failed component during refueling

outage.

32 ADVANTAGES AND DISADVANTAGES OF MECHANICAL VIBRATION

MONiTORING AND NEUTRON NOISE ANALYSIS

The preceding text describes similarities between the use of vibration monitoring

and neutron noise analysis for monitoring the condition of reactor vessel internals. The

following paragraphs describe the application and the advantages and disadvantages of

each measurement technique.

Vibration monitoring uses mechanical motion sensors such as accelerometers,

displacement transducers, or strain gages. The techniques for mounting the sensors and

conditioning their signals are well known, since these sensor types have been used for

many years in a variety of applications including rotating machinery monitoring, modal

and seismic testing, and predictive maintenance programs.' The success of vibration

monitoring depends on mechanical coupling between components to transfer vibrations

from the internal structures to the exterior of the reactor vessel or primary piping where

the sensors can be mounted.

Implementation of a vibration monitoring program in U.S. reactors would entail

installation of a number (between 4 and 12) of new sensors, thus posing a major hurdle

for establishing vibration monitoring programs in the United States because of the

understandable reluctance of utilities to make modifications inside reactor containment.

The main advantage of vibration monitoring is its direct nature: since vibration sensors

respond to a mechanical phenomenon that is mechanically transmitted to the sensors,

the measurements should be affected principally by changes in the mechanical conditions

within the system.
9

On the other hand, neutron noise analysis extracts and processes the time-varying

component of an existing neutron flux signal to gain indirect information concerning the

vibration of reactor vessel internals. This method is effective in detecting structural

vibration because as internal structures vibrate they induce small perturbations in the

reactor's neutron field. These perturbations are detected as fluctuations around'a

steady-state value in the neutron flux detector signals. The signal fluctuations (noise)

can be analyzed to obtain information about the nature of the perturbing vibrations.

The attractiveness of neutron noise analysis for U.S. utilities is that it uses existing

plant sensors, that is, the in-core and the ex-core neutron detectors that are part of

every reactor's instrumentation. Thus, implementation by U.S. utilities of vessel internals

monitoring programs based on neutron noise analysis appears to be considerably easier

than implementation of vibration monitoring programs.' The main disadvantages of

neutron noise analysis are that (1) measurements can be affected by phenomena other

than vibration and (2) only information concerning the reactor vessel internals can be

obtained; components of the primary system outside the reactor vessel (e.g., steam

generators) cannot be monitored. These points are amplified in the following

paragraphs.

'Because neutron noise analysis relies on neutron flux perturbations to detect

internals vibration, other phenomena that may also affect the flux must be considered

when analyzing neutron noise data. For example, noise signals' are known to be

influenced by' the fuel loading pattern, moderator boron concentration, and fuel burnup,

making it necessary to understand the effects of these variables on the noise signals.4

Thus, frequent baseline signature collection may be necessary when using neutron noise

analysis for monitoring the condition of reactor intemals, at least for the first several

fuel cycles.

As already noted, neutron noise analysis necessarily is restricted to structures

located within the reactor vessel. However, since components in the primary system are

coupled both mechanically and hydraulically, additional sensors (i.e., thermocouples,

pressure transducers, and accelerometers) located throughout the primary system may

enable the analyst to determine the cause of internals vibrations produced by sources

outside the reactor vessel. Obviously, the additional sensors provide information about a

larger structural system, thus making the monitoring program more complex but more

complete and potentially more useful.

In summary, both neutron noise analysis and vibration monitoring can be used for

monitoring the vibration of reactor vessel internal structures. Neutron noise analysis

would be easier to implement by U.S. utilities since the needed sensors are already in

place, but the information gained through its application is necessarily limited to

structures within the reactor vessel. Neutron noise data also are affected by phenomena

other than vibration, thus complicating the analysis and subsequent interpretation.

Vibration monitoring, although more difficult to implement because of the need for new

sensors, provides a more direct measurement of vessel internals vibration and allows a

monitoring program to be expanded, if desired, to include the entire primary coolant

system. Both methods employ essentially the same signal conditioning and data analysis

techniques and require the same amount of periodic data collection. Obviously, the

methods are not mutually exclusive. A case can be made for the combined use of

vibration monitoring and neutron noise analysis, an approach that has been widely

adopted in Europe.
4. SUMMARY OF OPERATING EXPERIENCE

This section summarizes the application of vibration monitoring and neutron noise

analysis for monitoring the condition of reactor vessel internal structures in both the

United States and Europe. The U.S. experience is presented first, followed by reported

experience in the Federal Republic of Germany (FRG) and France. Other countries

such as Sweden, Japan, Belgium, Hungary, and Finland have nuclear power plants in

which vibration monitoring or neutron noise analysis are used for monitoring the

condition of the vessel internals. However, the experience from these countries is not

included in this report simply because it is not as extensive or as widely reported as that

of the German and French programs.

4.1 U.S. EXPERIENCE

Surveillance of reactor internals vibration has seen only limited application in the

United States. Maintenance programs at U.S. utilities are typically composed of a

combination of corrective, preventive, and predictive maintenance techniques. Utilities

rely on generous design margins coupled with periodic in-service visual inspections to

ensure that the structural integrity of the internals will be maintained throughout the

service life of the reactor; they have not shown strong interest in real-time (but

somewhat indirect) means of monitoring internals condition.

A few utilities, however, have initiated vessel internals surveillance programs based

on neutron noise analysis and have demonstrated an ability to monitor vessel internals

vibrations." 7 Of these programs, one was cut back severely, one was eliminated, and one

was ineffectively implemented thereby allowing severe thermal shield damage to occur

even though progressive structural degradation clearly was indicated by the changing

noise signatures.8 9 It appears that, on the whole, the managements of these utilities did

not place sufficient importance on internals condition monitoring to provide the longterm support necessary for these programs to yield real benefits.

We are aware that internals monitoring programs are currently being performed by

Omaha Public Power District and by Northeast Utilities. In both programs, the utility

performs data collection activities and relies on the reactor supplier (in this case

Combustion Engineering) for support in data analysis and interpretation. This appears

to be a practical arrangement for utilities in the United States, where most of the

information and expertise required to perform interpretation of noise and vibration

signatures is proprietary and resides with the reactor supplier. This type of program, if

properly implemented, has the potential to produce valuable information on the

condition of the reactor vessel internals.

A more common application of reactor internals monitoring in the United States

involves relying on noise analysis to monitor, as a condition for continued plant

operation, the status of a degraded condition already known to exist. A recent proposal

by Southern California Edison to USNRC requesting continued operation of the San

Onofre Nuclear Generating Station 1 (SONGS 1) is an example of this type of internals

monitoring.'0 SONGS 1 has experienced degradation of thermal shield supports, so as a

condition of continued operation Southern California Edison proposed to monitor the

10
11

thermal shield vibration periodically using noise analysis of signals from the ex-core

neutron detectors to ensure that further degradation is not occurring.

An earlier but similar example involved instrument tube vibration in boiling-water

reactors (BWRs).Y In this case, neutron noise analysis was used to detect impacting of

instrument tubes on surrounding fuel channel boxes. By reducing the coolant flow rate

and reactor power, damaging impacts were eliminated, and the affected reactors could be

operated until a more satisfactory solution could be implemented. Noise monitoring was

used during this interim period to ensure that no further impacting occurred and also

was used after the permanent "fix" was installed to confirm that impacting indeed had

been eliminated.

U.S. research efforts in the application of noise and vibration analysis techniques

to commercial reactors have declined markedly in recent years. In the late 1970s and

early 1980s, researchers at ORNL studied and developed various aspects of noise

analysis including: BWR stability monitoring; automated noise data acquisition,

screening, and storage systems; core barrel and thermal shield vibration monitoring; and

the establishment of a baseline noise data signature library for all types of reactors using

data collected at a variety of U.S. power plants. This work represented state-of-the-art

reactor noise analysis at that time, and the 'results of this research were quickly

incorporated in vessel internals surveillance programs conducted abroad. In contrast to

these previous, extensive research efforts, almost no reactor-oriented noise analysis work

(with the exception of BWR stability monitoring) is being performed at ORNL today,

and the same can be said of other organizations that once were centers of U.S. activity

and expertise in this area. As support for this work continues to decline, research

capabilities are being lost through disuse, transfer of trained personnel, and rapidly

growing obsolescence and unreliability of data acquisition and analysis equipment.

4.2 GERMAN EXPERIENCE

In the FRG, neutron noise analysis and vibration monitoring of reactor internals

are a regular part of plant maintenance operations. German utilities monitor both

internals vibration and reactor coolant pump (RCP) shaft vibration on a continuing

basis. The sensor types and locations for a typical four-loop pressurized water reactor

(PWR) are shown in Fig. 2. The sensors consist of:

a 4 inductive absolute displacement sensors on the reactor vessel head flange

(A-signals);

* 16 relative displacement sensors, 2 below each RCP, measuring in 2 directions

and 2 near each steam generator, measuring horizontal and vertical

displacement (R-signals);

* 8 ex-core neutron detectors (X-signals);

* 5 piezoelectric pressure sensors at the inlet and outlet pipes (P-signals); and

* 8 proximity (absolute displacement) probes, 2 on each RCP, measuring pump

shaft displacement (W-signals)."2
12

* absolute displacement sensor

0 relative displacement sensor

(i piezo pressure sensor

I excore neutron noise detector

TJeddy current sensor

Fig. 2. Vibration and nois sensor positkm of a four-loop Gcrmnan PWR. (Source: R. Sunder,

Gesellschaft for Reaktorsicherheit. Reprinted by permission.)

The vibration signals are reviewed daily by plant personnel, who are thoroughly

trained in noise analysis techniques. Noise experts from Gesellschaft fur

Reaktorsicherheit (GRS), a private research organization similar to a national laboratory

in the United States, are consulted when deviations from baseline signatures or abnormal

trends are detected, and they also participate in the ensuing investigations until a

satisfactory conclusion is reached.
13

GRS has been involved in research programs aimed at improving methods for

monitoring nuclear reactor components for over 15 years. These programs involve not

only the reactor internals but also other parts of the primary system such as the RCPs,

piping, piping supports, and steam generators. GRS also evaluates vibration

measurements made periodically as part of the plant maintenance programs at German

pWRS.2

Previous work at GRS has resulted in the solution of several vessel internals

vibration problems. The most widely reported example involves using vibration

monitoring to detect the relaxation of core support barrel hold-down springs."' 4 In this

case, a frequency shift in the resonance associated with the upper core barrel and core

structure was observed. Results from an analytical model indicated that the most likely

cause of this shift was relaxation of the core support barrel hold-down springs. New

springs were ordered while the internals vibration signatures were monitored closely for

signs of additional changes in structural integrity throughout the remainder of the fuel

cycle. Following vessel head removal for refueling, the hold-down springs were

examined, and 65 of 112 produced insufficient holding force to meet specification. The

new springs were installed, and the internals vibration signatures immediately returned to

normal. In this incident, a vibration monitoring program coupled with analytical model

results allowed an anomaly to be detected and diagnosed and the reactor to continue

operation for the remainder of the fuel cycle. Sufficient time was made available for

replacement parts to be ordered, resulting in correction of the problem with minimal

impact on plant operation.

More recently, GRS has been involved in monitoring for loss of preload in screws

holding core former plates to their baffles. Loosening of these screws may give rise to

loose parts circulating in the reactor coolant system and has caused an extended outage

at one German PWR.2 GRS researchers believe this particular degradation can also be

detected by their on-line monitoring systems.

GRS has developed and tested a system for use by trained plant personnel and

operators that continuously monitors vibration of the RCPs and the reactor internals. A

prototype of the system, named COMOS (COndition MOnitoring System), was tested on

the Grafenrheinfeld pressurized water reactor for nearly one full fuel cycle, and updated

versions of COMOS are being installed in the newest Kraftwerk Union (KWU)

'Convoy" plants.2 Both frequency and amplitude of resonances are monitored to detect

deviations from reference (baseline) spectra. Abnormal deviations are detected by

comparing a quotient spectrum, which is calculated from current and reference spectra,

to preset thresholds. The system produces no control-board alarms, but the plant

operations staff reviews the analysis results daily. Abnormal spectra are referred to GRS

personnel for analysis and interpretation based on results from analytical model studies.

A 45-DOF finite-element model of the primary system (including the reactor vessel,

internal structures, steam generator, and RCP) has been developed and validated against

measured vibration data.' This model is shown in simplified form in Fig. 3.

Five beams are used to represent the reactor pressure vessel (RPV) and its

internals; referring to Fig. 3(b), these beams represent (1) RPV, (2) core barrel,

(3) lower support structure, (4) upper plenum assembly, and (5) fuel. Figure 3(b) also

shows the pinned connections joining the beam elements. Each connection has a

rotational stiffness, and each beam has an associated mass distribution and flexural

stiffness. The model calculates pendular, vertical, and rotational displacements of the

internals.
14

Steam generator (SG)

Main coolant pump (MCP)

Reactor pressure vessel

(a) PWR prmary arcuiL

Fig. 3 Modeling of PWR primar aryiit and reactor internal (Source: V. Bauernfeind,

Gesellschaft fur Reaktorsicherheit. Reprinted by permission.)

The remainder of the model represents one loop of the primary system. Again,

beam elements are used to model these components, allowing calculation of translation,

rotation, and bending displacements. The piping and component supports are modeled

using springs. In addition to natural frequency and mode calculations, the model can be

used to calculate the response to applied forces, such as flow excitation. Component
15

(b) Rigid body nodcl rprezitation of RPV itnas

Fig 3. (cotinued).

masses were obtained from manufacturer's data, and results from shaker tests were used

to estimate the model rotational and flexural stiffness, which then were adjusted until

model results closely matched measured data. After validating the model, the results can

be used to study the effects of various postulated forms of mechanical degradation on

internals vibration. This model was used in the previously mentioned example of relaxed

core barrel hold-down springs to postulate the probable cause of the observed deviation

from the baseline vibration signatures. It also has been used to determine the alarm

criteria built into the COMOS system.' Thus, this model is a valuable tool used

extensively by GRS to gain understanding of how the core internals are expected to

behave when subjected to various postulated degradations. Within the limitations of the

model, the understanding thus gained allows abnormal vibrations to be diagnosed and

appropriate actions taken.

The German program for monitoring vibrations of primary system components,

including the reactor internals, is extensive and well implemented. It has evolved to the

point that vibration and neutron noise data are continuously monitored by automated

systems, with daily review by plant personnel being a normal part of plant maintenance

activities. The development and validation of analytical models are a keystone in this

program, allowing GRS personnel to accurately interpret abnormal noise signatures so

that the seriousness of any anomaly may be ascertained and appropriate actions taken.
16

43 FRENCH EXPERIENCE

A similar situation regarding noise and vibration analysis exists in France. Reactor

internals condition monitoring is performed for all French plants by the Surveillance and

Diagnostic Maintenance Group at Electricitt de France (EDF). This group is

responsible for research, development, application, and training in the areas of reactor

maintenance, surveillance of reactor internals and rotating machinery, and diagnosis of

abnormal vibration and loose parts. EDF operates a comprehensive monitoring

program, like that in the FRG, that includes regular data collection and analysis as a

part of normal plant maintenance practices.

Signals used to monitor primary system components in French PWRs are obtained

from accelerometers mounted on the reactor vessel, steam generators, primary system

piping, and pump casings, as well as from pump shaft displacement transducers and noise

signals from the in-core and ex-core neutron flux detectors.2' 6 These sensor locations as

well as a block diagram of the entire monitoring system are shown in Fig. 4. The

accelerometers share the dual roles of monitoring vibration and detecting loose parts.

Each month, signals are collected from the ex-core detectors and from accelerometers

mounted on the reactor vessel and analyzed using a spectrum analyzer and a

microcomputer. The computer compares the collected signatures with baseline

signatures acquired at the beginning of the fuel cycle and thus determines whether or

not an anomaly is present. A more extensive analysis of all signals is performed every

3 months. The vibration data are sent to EDF, where they are transferred to a central

data base. This data base includes vibration signatures obtained during hot functional

and shaker tests, as well as 80 fuel cycles of operating history covering 45 reactor units

(as of May 1987).2

In contrast to the German programs, interpretation of anomalous signatures is

performed at EDF headquarters in Paris rather than at the plant sites. Interpretations

are based on modeling results, results from hot functional tests, the vibration and

neutron noise signatures collected periodically and stored in the data base, and

experience with previously encountered anomalies.

An additional unique tool available to EDF for model verification and anomaly

diagnosis is the SAFRAN test loop located in Saclay."7 SAFRAN (Fig. 5) is a 1/8th

scale mock-up of the primary system of a 900 MW(e) French PWR. SAFRAN includes

a detailed representation of the reactor vessel complete with internals; pressurizer; and

the piping, pumps, and steam generators present in each loop. The mock-up is

instrumented with 80 sensors, including pressure transducers, displacement probes to

measure core barrel and thermal shield movement, and accelerometers. SAFRAN has

been used to estimate damping rates for use in analytical models, to validate the model

results, and to investigate directly the effect of various degradations on vibration

signatures. The heavily instrumented internals allow modeshapes to be verified

experimentally, a capability unavailable to other researchers.

Like the Germans, the French have been able to detect and diagnose abnormal

internals vibration using surveillance and monitoring techniques. Two examples are

described subsequently: the first involves detection of baffle jetting in 1981-82, and the

second is a more recent occurrence involving excessive instrument tube vibration.

Baffle jetting is a phenomenon that can occur because of the gaps between the

baffle plates that separate flow in the coolant downcomer from upward core flow. A

pressure difference exists between the coolant on each side of the baffle; thus, the
17

LOOP 3

LOOP I

AT2

CN NEUTRONIC EX

AT: ACCELEROMETER

AG: ACCELEROMETER

CORE CHAMBER

ON TUBE GUIDE

ON STEAM GENERATOR

(a) Vi oatmo and nDosc sensor p tmiu

CONTROL ROOM INDICATION

PRIMARY = OIITININ AND ALARM PUMPTONN -

.VELOCITY,

DISPERCEMEN... CASSEi

SEWSORS RECORI

I CONTROL ROom dDICATION j

I - AND ALARM i fij

(b) Bloek diagram of mncitainn ystem

FI& 4. Monitdg syst= em intl d PWRs cquippcd vnth parta iermal shiekl (Source: C Puyal,

Electricite de France. Reprinted by permission.)
18

Fig. 5. SAFRAN test loop. (Source: J. C. Carre, Commissariat a L'Unergie Atomique. Reprinted by permission.)

presence of an unintentional gap causes a horizontal coolant "jet' to impinge against an adjacent outer fuel assembly. Baffle-jetting damage was first encountered in France in 1981 on the Bugey 2 reactor during its second fuel cycle." A sudden increase in primary water activity was observed, which was traced to a cladding breach in a peripheral assembly. Two damaged fuel assemblies were found at the end of the cycle. In response to this finding, a monitoring program based on in-core neutron noise signals was initiated by EDF in Bugey and Fessenheim (another French PWR of similar design). The measurements showed a clear anomaly, namely excessive spectral energy between 20 and 30 Hz. 6 The source of this anomaly appeared to be an assembly near
19

the baffle jet in Bugey Unit 2. Results of the monitoring program allowed several

conclusions to be drawn. First, baffle jetting may not always be detectable, even in

assemblies directly in the jet's path, if the outermost fuel pins on which the jet impinges

do not transmit their vibration to the remainder of the fuel assembly. Second, baffle

jetting may sometimes cause a fuel assembly adjacent to the one actually being impinged

upon to vibrate, thereby generating excessive neutron noise. Finally, in any event the

vibration signature anomaly is detectable only by sensors positioned close to the jet

location. Thus, noise signatures from in-core detectors are probably necessary to detect

the presence of baffle jetting and thereby allow appropriate maintenance to be

scheduled during the next refueling.

A more recent problem experienced in many 1300 MW(e) French plants is

excessive instrument tube vibration caused by flow turbulence below the core support

plate.2', These vibrations eventually lead to perforation of the tube at or above the

core support plate, allowing cold water to impinge on adjacent fuel rods and thereby

cause cladding failure. Several methods of stiffening the connection between the

instrument thimble and the core support plate have been used to reduce the vibrations.

The effectiveness of stiffening the instrument-tube/core-support-plate connection was

verified by analyzing vibration data from accelerometers mounted on the tubes. The

experience gained from this investigation allowed a new approach to this problem.

Accelerometers are now mounted on all instrument tubes and their vibration is

monitored daily. When incipient failure of a tube is predicted, the tube is replaced as

soon as possible.

The French, like the Germans, have implemented a comprehensive and effective

program for monitoring the components in the PWR primary coolant systems. Vibration

and neutron noise data are monitored as a normal part of plant maintenance, enabling

component degradation to be detected and providing additional information that can be

used to identify the degraded component.

4.4 COST EFECTIVENES

It is evident from these examples and others reported in the literature that

internals condition monitoring and analysis research programs can provide valuable

information to utilities and licensing authorities on nuclear plant operability and

condition. Reliable estimates of benefit-to-cost ratios are not so easily found, however.

EDF has recently estimated the benefit-to-cost ratio of their monitoring and noise

research programs to be -2:1.'9 This figure is based on actual costs and the estimated

value of savings accumulated between 1976 and 1986 attributable to the reactor noise

and vibration monitoring programs. Since these EDF programs include vibration,

neutron noise, and loose parts monitoring, the cited 2:1 benefit-to-cost ratio reflects the

combination of these three activities, for which a separation has not been provided. The

cumulative cost is the total spent on design and development, equipment, personnel

training, and routine data collection and analysis. Benefits were calculated using

estimates of the damage that would have occurred, had the monitoring system not been

present. Damage, repair, and outage estimates were based on similar incidents that had

occurred at other EDF facilities, and the financial benefits include the estimated repair

and outage costs.
20

Because of the similarity in plant and equipment design between European and

U.S. plants, it seems reasonable to expect that reactor vessel internals condition

monitoring programs applied in the United States would yield benefits comparable to

those experienced abroad. Based on European experience and an emerging consensus

among U.S. noise analysts, guidelines have been developed to assist U.S. utilities in

initiating and executing an internals monitoring program.' These guidelines are

reviewed in Sect. 5.
S. UTILiTY GUIDELINES FOR THE IMPLEMIENTATION OF INTERNAIS

MONITORING PROGRAMS

This section describes some aspects of the implementation of an internals

condition monitoring/neutron noise analysis program by U.S. utilities. It is recognized

that although a well-designed and well-implemented internals monitoring program can be

an asset to plant operation and safety, a poorly organized or poorly implemented

program can actually be detrimental to plant operation, producing low-quality data that

may lead to false alarms under normal conditions or lack of alarms under abnormal -

conditions. Some important considerations of noise monitoring program implementation

are: (a) program organization, (b) integration of the program into plant or utility

operations, (c) formulation of data acquisition and analysis procedures, and (d) personnel

training in noise and vibration analysis techniques.' This section (5) closely follows

Sect. 3 of ref. 20.

Reactor internals condition monitoring/neutron noise analysis programs are

typically organized at the utility engineering level or plant operations level, or they are

dependent on expertise supplied by outside consultants. Table 3 lists some of the

advantages and disadvantages inherent in each type of organization.

Organization at the utility engineering level is most advantageous for utilities

having multiple reactor units or when personnel or equipment is limited. This level of

organization reduces duplication of equipment and personnel positions, but

communication between plant personnel and the noise/vibration analyst may be

compromised. If permanent access to plant signals is not provided, data acquisition and

quality assurance (QA) activities may become costly and inefficient.

If a utility has only one plant or can afford to support programs at each plant site,

then organization at the plant level may be advantageous. Data acquisition, QA, and

integration of the program into plant operations are thereby simplified. However,

communication between noise/vibration analysts at different plants and between the plant

programs and design or analysis groups at the utility engineering level must be ensured.

This is especially important if a utility operates several units of similar design.

Specialist consultants are useful when plant personnel are unable (or untrained) to

solve a particular noise analysis problem. It is nonetheless desirable for the utility to

maintain its own in-house expertise to evaluate the results and understand the

interpretations and recommendations of the consultants. The utility must also take steps

to ensure that adequate information is provided to the consultant so that he or she gains

a proper appreciation of relevant plant history as well as the current problem.

It should be pointed out that the levels of organization just described may be

applied to different aspects of the same program. The German and French programs

provide examples of this. In both the FRG and France, data collection, screening, and

integration with plant activities are performed at the plant leveL With EDF, central

data storage, analysis, and research are all performed at the utility level, whereas in the

FRG, GRS provides the analysis and research capability. The monitoring programs in

place at Northeast Utilities and Omaha Public Power District both carry out plant noise

analysis programs, and both have service arrangements with Combustion Engineering,

which provides diagnostic and analysis services.

21
22

Table 3. Advantages and disadvantages of reactor nowe program organiationz

Organization Possible Possible

level advantages disadvantages

Utility Fewer personnel Manpower shortages

Less equipment duplication Lack of:

Communication with design groups plant access

Pooling of analysis results signal access

communication with plant

Plant Knowledge of plant activities More training required

Knowledge of plant instrumentation More personnel required

Signal access Equipment duplication

Personnel availability Lack of:

communication among plants

communication with utility

communication with design

and analysis groups

Consultant May be called when needed Limited problem solution

May supply own equipment Knowledge leaves with

Fast problem solution consultant

Knowledge and experience Lack of:

Low short-term costs results verification

plant access

knowledge of plant

activities

communication with plant

or utility

The effectiveness of a vessel internals condition monitoring/neutron noise analysis

program depends to a large extent on how well the program is integrated into overall

plant operations. Communication between noise/vibration analysts and the plant is vital

if the analyst is to know about operational changes or problems that may have an effect

on noise signatures. Likewise, the plant staff must be aware of abnormal noise

signatures and their interpretations if corrective action is to be taken. Good

communication usually requires that the noise/vibration analyst and plant personnel

jointly review results at least quarterly (preferably monthly).

Procedures must be developed to ensure that noise data acquisition and analysis

are carried out properly and are effectively integrated with plant activities. Procedure

formats are usually dictated by utility guidelines. Well-formulated procedures for data

acquisition and analysis will include the purpose of the procedure, execution time and

personnel requirements, precautions, limitations and actions, prerequisite plant

conditions, required equipment, data to be collected, acceptance criteria, and any special

instructions.

Highly trained noise/vibration analysts with extensive analysis experience are.

essential to the success of an internals condition monitoring/neutron noise analysis

program. Based on experience at ORNL, the subjects the noise/vibration analyst should
23

be familiar with are listed in Table 4. Although the analyst should assume responsibility

for all aspects of the internals condition monitoring program, it is possible to have

certain aspects of the program such as data collection and reduction performed by

personnel not trained specifically in noise or vibration analysis.

Table 4. Desed knwl Fr reactor Dos anaLy D

Random processes

Random signal analysis

Fourier transforms

Statistics

Nuclear plant instrumentation

Digital signal processing

Fast Fourier transform (FF1I) algorithms

Use of spectrum analyzers

Use of tape recorders

Fluid flow

Heat transfer

Reactor dynamics

Structural analysis

Electronics

State-of-the-art noise analysis
6 CONCLUSIONS

Vibration monitoring/neutron noise analysis has progressed over the past 20 years

from a research topic to a practical tool that can be (and is) used to obtain information

on the condition of a nuclear plant's reactor vessel internals and other primary coolant

system components. Monitoring programs implemented in both France and the FRG

have demonstrated the benefits of these methods, and these programs can be used by

U.S. utilities as models for constructing their own reactor vessel internals condition

monitoring programs. The success of such programs in the United States would depend

on the priority utilities give them. If the guidelines outlined above and discussed in

greater detail in ref. 20 are followed during the creation and operation of a reactor

vessel internals condition monitoring program, one can expect that economic and safety

benefits similar to those obtained by the French and German programs will be realized.

This would give U.S. utilities a tool with which to improve daily plant operation, monitor

this one aspect of plant structural aging, perform more effective maintenance, and

provide a portion of the technical basis necessary for evaluating the feasibility of

extended plant operation without replacement of reactor vessel internal structures.

24
ACKNOWLEDGMENT

The authors would like to thank the Electric Power Research Institute and

individuals representing U.S. reactor manufacturers and nuclear utilities for their valuable

constructive criticism of a draft version of this report

25
REFERENCES

1. J. S. Mitchell, An Introduction to Machinety Analysis and Monitoring, Pennwell

Publishing Co., Tulsa, Okla., 1981.

2. F. J. Sweeney, Oak Ridge National Laboratory Foreign Trip Report,

ORNL'FIR-2537, Martin Marietta Energy Systems, Inc., Oak Ridge National

Laboratory, April 1987.

3. B. Damiano, Reactor Internals Surveillance for Plant Life Extension, U.S.

Nuclear Regulatory Commission informal letter report ORNL/NRC/LTR-86/11, Martin

Marietta Energy Systems, Inc., Oak Ridge National Laboratory, April 1986.

4. F. J. Sweeney and J. A. Renier, Sensitivity of Detecting In-Core Vibrations and

Boiling in Pressurized Water Reactors using Ex-Core Neutron Detectors, U.S. Nuclear

Regulatory Commission report NUREG/CR-2996, Oak Ridge National Laboratory, 1984.

5. G. P. Horne, "Recent Experience with Noise Analysis at Oconee Nuclear

Station," Prog. Nucl. Energy 15, 395-402 (1984).

6. R. S. Johnson, Neutron Noise Analysis Program for Surveillance of Reactor

Internals Vibration, TR-NF-1022, Duke Power Company, Charlotte, N.C., 1985.

7. J. W. Quinn, J. A. Stevens, and C. J. Sterba, "The Use of Ex-Core Neutron

Noise at Near-Zero Reactor Power to Monitor Thermal Shield Support System

Integrity," Prog. NucL Energy 21, 127-33 (1987).

8. F. J. Sweeney and D. N. Fry, "Thermal Shield Support Degradation in

Pressurized Water Reactors," Flow-Indtuced Vibration - 1986, PVP-Vol. 104, American

Society of Mechanical Engineers, New York, 1986.

9. B. T. Lubin, R. Longo, and T. Hammel, "Analysis of Internals Vibration

Monitoring and Loose Part Monitoring System Data Related to the St. Lucie 1 Thermal

Shield Failure," Prog. Nucl. Energy 21, 117-26 (1987).

10. F. R. Nandy, "Concerns Regarding Thermal Shield at SONGS 1," letter from

Southern California Edison Co. to USNRC, Mar. 21, 1989, Enclosure 1.

11. D. N. Fry, R. C. Kryter, and M. V. Mathis, Summary of ORNL Investigations

of In-Core Vibrations in BWR-4s, ORNL/NUREG/TM-101, Oak Ridge National

Laboratory, 1977.

12. R. Sunder, "On-Line Monitoring of Neutron Noise and Vibration Signatures;

Operating Experience with COMOS in German PWRs," 1988 Informal Meeting on

Reactor Diagnostics, Orlando, Fla., June 8-10.

26
27

13. R. Sunder and D. Wach, "Reactor Diagnosis using Vibration and Noise

Analysis in PWRs," in Operational Safety of Nuclear Power Plants I International Atomic

Energy Agency, Vienna (1984).

14. W. Basti, R. Sunder, and D. Wach, 'The Influence of Noise Diagnostic

Techniques on the Safety and Availability of Nuclear Power Plants," Prog. Nucl Energy

15, 513-24 (1984).

15. V. Bauernfeind, 'Vibration Monitoring of a Four-Loop PWR: Model

Investigations of the Sensitivity of the Monitored Signals on Mechanical Failures," Prog.

NucL Energy 21, 247-54 (1987).

16. C. Puyal, A. Fernandes, and C. Vincent, "Primary System Surveillance and

Diagnostics of PWR Power Plants in France," 5th Power Plant Dynamics, Control, and

Testing Symposium, Knoxville, Tenn., Mar. 21-23, 1983.

17. J. C. Carre et al., "PWR Internals Vibrational Mode Shapes Calculation and

Tests," Prog. Nucl. Energy 1, 353-62 (1977).

18. C. Puyal et al., "Solution to Thimble Vibrations of French and Belgian

Reactors using Accelerometers and Neutron Noise," 19th Informal Meeting on Reactor

Noise, Rome, June 4-6 1986.

19. B. Michel and C. Puyal, "Operational and Economical Experience with

Vibration and Loose Parts Monitoring Systems on Primary Circuits in PWRs," Prog.

Nucl Energy 21, 469-74 (1987).

20. F. J. Sweeney, Utility Guidelines for Reactor Noise Analysis, EPRI N13-4970,

Electric Power Research Institute, Palo Alto, Calif., 1987.
29

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2. TITLE AND SUBTITLE

Current Applications of Vibration Monitoring and Neutron Noise

Analysis: 3. DATE REPORT PUBLISHED

Detection and Analysis of Structural Degradation of Reactor February 1990

Vessel Internals from Operational Aging 4. FIN OR GRANT NUMBER

______________________________________ B0828

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B. Damiano, R.C. Kryter

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10. SUPPLEMENTARY NOTES

11. ABSTRACT 1200 words or less;

This report, which was prepared under the Nuclear Plant Aging Research Program

sponsored by the United States Nuclear Regulatory Commission, discusses the

application of vibration monitoring and neutron noise analysis for monitoring

light water reactor (LWR) vessel internals. The report begins by describing

the effects of loss of structural integrity on internals vibration and how

sensible parameters can be used to detect and track the progress of

degradation. This is followed by a description and comparison of vibration

monitoring and neutron noise analysis, two methods for monitoring themechanical integrity of reactor vessel internal components. The major section

of the report describes the status of reactor vessel internals condition

monitoring programs in the United States, Federal Republic of Germany, and

France, three countries having substantial commitments to nuclear power. The

last section presents guidelines for U.S. utilities wishing to establish

reactor internals condition monitoring programs.

12. KEY WORDS/DESCR!PTORS (List words orphrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

reactor vessel internals unlimited

vibration 14. SECURITY CLASSIFICATION

neutron noise analysis (This Page)

structural degradation unolassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-891
T , I t:

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

IPOSTAGI & FEES PAIDI UUSNRC

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