U.S. patent number 4,961,898 [Application Number 07/238,051] was granted by the patent office on 1990-10-09 for reactor internals and core support monitoring system.
This patent grant is currently assigned to Westinghouse Electric Corp.. Invention is credited to William T. Bogard, George J. Bohm, William Ciaramitaro, Sam S. Palusamy, Norman R. Singleton, John A. Tgortorice.
United States Patent |
4,961,898 |
Bogard , et al. |
October 9, 1990 |
Reactor internals and core support monitoring system
Abstract
A nuclear reactor monitoring system includes incore and excore
neutron detectors for producing neutron noise signals and system
data including at least one of reactor coolant pump motor current,
temperature and pressure of the reactor coolant system and
pressurizer pressure and means for diagnosing conditions of the
reactor internals and core support structures by comparing analyzed
instrumentation signatures. The signatures may be derived from
scale model test data, plant test data, analytical model results
and trending analysis results. Analyses performed by the system
include time and frequency domain analyses of the neutron noise
signals and the system data. The analyzed data are compared with
rules in a rules base of an expert system to generate a diagnosis
of the condition of the reactor internals and core support
structures and action recomendations for maintenance of the nuclear
reactor.
Inventors: |
Bogard; William T. (Irwin,
PA), Bohm; George J. (Pittsburgh, PA), Ciaramitaro;
William (Murrysville, PA), Palusamy; Sam S.
(Murrysville, PA), Singleton; Norman R. (Pittsburgh, PA),
Tgortorice; John A. (Export, PA) |
Assignee: |
Westinghouse Electric Corp.
(Pittsburgh, PA)
|
Family
ID: |
22896300 |
Appl.
No.: |
07/238,051 |
Filed: |
August 30, 1988 |
Current U.S.
Class: |
376/245;
976/DIG.207; 706/915 |
Current CPC
Class: |
G21C
17/00 (20130101); G21D 3/001 (20130101); Y02E
30/00 (20130101); Y10S 706/915 (20130101); Y02E
30/30 (20130101) |
Current International
Class: |
G21C
17/00 (20060101); G21C 017/00 () |
Field of
Search: |
;376/245,249,259,258
;73/579,590 |
References Cited
[Referenced By]
U.S. Patent Documents
|
|
|
3860480 |
January 1975 |
Carteus et al. |
3860481 |
January 1975 |
Gopal et al. |
4060716 |
November 1977 |
Pekrul et al. |
|
Primary Examiner: Wasil; Daniel D.
Attorney, Agent or Firm: Possessky; E. F.
Claims
What is claimed:
1. A method for diagnosing conditions of reactor internals and core
support structures in a nuclear reactor, comprising the steps
of:
(a) detecting vibration of reactor internals;
(b) receiving signals representing system data including at least
two among pressurizer pressure, reactor coolant pump motor current,
reactor coolant system pressure, reactor coolant system temperature
and reactor coolant system flow rate; and
(c) automatically evaluating the vibration of the reactor internals
detected in step (a) based upon all of the signals received in step
(b) to diagnose the condition of the reactor internals and core
support structures.
2. A method as recited in claim 1,
further comprising the step of (d) automatically evaluating the
vibration of the reactor internals based upon structural data
including at least one of displacement, acceleration and strain of
the core support structures, and
wherein step (c) comprises the steps of:
(ci) analyzing neutron activity emanating from the core of the
nuclear reactor using cepstrum analysis to produce an internal
vibration indicator;
(cii) analyzing the system data using cepstrum analysis on at least
one of the pressurizer pressure, reactor coolant pump motor
current, reactor coolant system pressure and displacement,
acceleration and strain of the core support structures to produce a
system indicator of reactor operation; and
(ciii) comparing the system and internal vibration indicators with
rules defining a relationship between structural abnormalities and
signatures of the system data and the neutron activity to produce a
diagnosis of the condition of the reactor internals and core
support structures.
3. A method as recited in claim 2, wherein step (cii) further
comprises applying analysis using Duhamel's theorem to the
temperature of the reactor coolant system in producing the system
indicator of reactor operation.
4. A method as recited in claim 1, wherein the signals received in
step (b) represent the system data including at least the
pressurizer pressure, the reactor coolant pump motor current, the
reactor coolant system pressure, the reactor coolant system
temperature and the reactor coolant system flow rate.
5. A system for diagnosing conditions of reactor internals and core
support structures in a nuclear reactor, comprising:
monitoring means for detecting vibration of reactor internals and
for obtaining system data by detecting at least two of pressurizer
pressure, reactor coolant pump motor current, reactor coolant
system pressure, reactor coolant system temperature and reactor
coolant system flow rate; and
diagnosis means for automatically evaluating the vibration of the
reactor internals based upon the system data to diagnose the
condition of the reactor internals, and the core support
structures.
6. A system as recited in claim 5, wherein said monitoring means
comprises:
neutron detectors for detecting neutron activity emanating from the
core of the nuclear reactor; and
at least two of:
a current detector for detecting the reactor coolant pump motor
current;
a temperature detector for detecting the reactor coolant system
temperature; and
pressure detectors for detecting the pressurizer pressure and the
reactor coolant system pressure.
7. A system for diagnosing conditions of reactor internals and core
support structures in a pressurized water nuclear reactor having a
reactor vessel and a reactor coolant system with a temperature and
pressure maintained by a reactor coolant pump motor, said system
comprising:
neutron detectors for outputting a first signal indicative of
neutron activity emanating from the core of the nuclear
reactor;
a current detector for outputting a second signal indicative of
current drawn by the reactor coolant pump motor;
a temperature detector for outputting a third signal indicative of
the temperature of the reactor coolant system;
pressure detectors for outputting fourth and fifth signals
indicative of pressurizer pressure and the pressure of the reactor
coolant system;
signal communication and isolation means for transmitting the
first, second, third, fourth and fifth signals; and
a reactor vessel diagnostic unit, connected to said signal
communication and isolation means, for analyzing the neutron
activity, the temperature, pressure and flow rate of the reactor
coolant system, the pressure of the pressurizer and the current of
the reactor coolant pump motor and for diagnosing the condition of
the reactor internals and core support structures by comparing the
results from the analysis with predetermined system data.
8. A system as recited in claim 7, further comprising at least one
of:
a flowmeter for outputting a sixth signal indicative of the flow
rate of the reactor coolant system;
an accelerometer for measuring motion of at least one of the
reactor vessel, reactor vessel internals and a dynamically coupled
reactor coolant system component to produce a seventh signal;
a displacement detector for measuring displacement of at least one
of the reactor vessel, the reactor vessel internals and the
dynamically coupled reactor coolant system component to produce an
eighth signal; and
a strain gauge for measuring strain in at least one of the reactor
vessel, reactor vessel internals and the dynamically coupled
reactor coolant system component to produce a ninth signal, and
wherein said signal communication and isolation means transmits and
said reactor vessel diagnostic unit analyzes at least one of the
sixth, seventh and eighth signals.
9. A system as recited in claim 8, wherein said reactor vessel
diagnostic unit includes means for producing a maintenance
scheduling recommendation in dependence upon the condition of the
reactor internals and core support structures.
10. A system as recited in claim 8, wherein said signal
communication and isolation means transmits and said reactor vessel
diagnostic unit analyzes unit analyzes the data represented by all
of the first through ninth signals.
11. A method for diagnosing conditions of reactor internals and
core support structures in a nuclear reactor having components,
comprising the steps of:
(a) receiving sensor data indicating operational states at least
two of the components of the nuclear reactor, the sensor data
including excore and incore neutron detector noise signals; and
(b) analyzing the sensor data in comparison with scale model test
data, full scale plant test data, analytical model results and
trending analysis results to diagnose the conditions of the reactor
internals and the core support structures.
12. A method as recited in claim 11, wherein step (b) comprises
applying expert system rules defined using signal signatures
derived from the test data and the analytical model and trending
analysis results to identify degradation mechanisms of the reactor
internals and the core support structures.
Description
BACKGROUND OF THE INVENTION
1. Field of the Invention
The present invention is related to monitoring of a nuclear reactor
and, more particularly, to a system for monitoring and diagnosing
abnormalities within a nuclear reactor pressure vessel.
2. Description of the Related Art
A variety of monitoring is conventionally performed on nuclear
reactors for early detection of degradations to minimize plant
outages. Examples of such monitors can be found in U.S. Pat. Nos.
3,860,480 and 3,860,481, both assigned to the assignee of this
invention. These patents are directed respectively to analysis of
the vibration of core components as indicated by analysis of the
noise content of neutron flux signals and a loose parts monitor
using accelerometers mounted on the exterior surfaces of primary
system components where metal particles are likely to accumulate.
Many other types of monitoring systems are used on a typical
nuclear reactor. Some systems include analysis of the signals,
while others simply consist of a sensor, a recorder and
communication means for supplying the signals detected by the
sensor to the recorder.
Thus, there is a great deal of information about the operation of
components and systems generated for a typical nuclear reactor.
However, there are relatively few monitoring systems which analyze
the available data to determine possible causes of changes in
signals. As a result, nuclear reactor plants used for, e.g.,
electricity generation, experience unplanned outages when
components fail or approach failure, despite the information
provided by existing monitoring systems.
SUMMARY OF THE INVENTION
An object of the present invention is to provide a monitoring
system for identifying signal signatures indicative of structural
abnormalities.
Another object of the present invention is to provide a monitoring
system for diagnosing system abnormalities prior to failure.
Yet another object of the present invention is to reduce the number
of unplanned outages.
A further object of the present invention is to provide maintenance
scheduling recommendations based upon the existence of signal
signatures in the signals detected by monitoring a nuclear
reactor.
A still further object of the present invention is to extend the
life of a nuclear reactor by appropriately scheduling
maintenance.
The above objects are attained by providing a method for diagnosing
conditions of reactor internals and core structures in a nuclear
reactor, comprising the steps of: detecting vibration of reactor
internals and automatically evaluating the vibration of the reactor
internals based upon system data including at least one of
pressurizer pressure, reactor coolant pump motor current, reactor
coolant system pressure, reactor coolant system temperature and
reactor coolant system flow rate, to diagnose the condition of the
reactor internals and core support structures. Structural data,
including at least one of displacement, acceleration and strain of
the reactor vessel or reactor internals, may also be used in
evaluating the reactor internal vibration. Analysis techniques for
evaluating the data include auto and cross power spectral density
functions, cepstrum analysis, Duhamel's theorem and other frequency
and time domain statistical properties.
Preferably, the data is evaluated using an expert system defining a
relationship between structural abnormalities and signatures of the
system data and the neutron activity to produce a diagnosis of the
condition of the reactor internals and core structures.
These objects, together with other objects and advantages which
will be subsequently apparent, reside in the details of
construction and operation as more fully hereinafter described and
claimed, reference being had to the accompanying drawings forming a
part hereof, wherein like reference numerals refer to like parts
throughout.
BRIEF DESCRIPTION OF THE DRAWINGS
FIG. 1 is a block diagram of a monitoring system according to the
present invention; and
FIG. 2 is a block diagram of an expert system used to perform
diagnostics in the monitoring system illustrated in FIG. 1.
DESCRIPTION OF THE PREFERRED EMBODIMENTS
As nuclear reactors used for electricity generation have become
older, there has been increasing interest with predicting failure
of components to extend the life of such plants and minimize
unplanned outages required for repairs. Deterioration in components
such as loss of integrity of thermal shield supports, cracking in
guide tube support pins, loss of preload in the core barrel hold
down spring, fuel rod wear induced by flow and influenced by
internal responses, upper internals flow induced vibration and
reactor core misalignment are examples of reactor internals
degradation which have been experienced. In some cases,
deterioration of this sort has lead to significant unplanned
outages. Although monitoring may not reduce or prevent
deterioration, useful information can be provided on recommended
modes of operation and timing of repair and replacement of
components. The reduction of unplanned outages can provide
significant savings, several times the cost of a monitoring system
according to the present invention.
These savings are possible by the use of detectors currently in
place at most nuclear power plants along with additional
instrumentation if justified. A typical plant includes excore 10
and incore 11 neutron detectors, a reactor coolant pump motor
current detector 12 and temperature detectors 14. Usually, pressure
sensors 15, 16 also exist at nuclear power plants to measure
pressurizer pressure 15 and reactor coolant system pressure 16. In
addition, a loose parts monitoring system 18 like that described in
U.S. Pat. No. 3,860,481 is commonly found at nuclear power
plants.
However, the sensors used in these systems are simply connected to
recorders or, as described in U.S. Pat. Nos. 3,860,480 and
3,860,481 used only to evaluate the data from a single type of
sensor. According to the present invention, signals from the excore
and incore neutron detectors 10, 11 are supplied together with at
least one of the other sensors via signal communication and
isolation means 20 to a reactor vessel vibration diagnostics unit
22. The signal communication and isolation means 20 may be provided
by conventional cabling or the system described in U.S. application
Ser. No. 934,238 filed Nov. 20 1987, now U.S. Pat. No. 4,770,842
and conventional conditioning and isolation circuitry. The reactor
vessel diagnostics unit 22 outputs the results of analysis
performed as described below on conventional output devices 24,
such as computer displays, printers, plotters, etc.
In addition to the detectors which are currently installed at most
operating nuclear power plants, other transducers can be used to
provide additional information to help diagnose specific
conditions. For example, rugged, high sensitivity accelerometers 26
and fluctuating pressure transducers 28 can be added to provide
information about small or quick vibrations and pressure
fluctuations. Other special transducers 30, such as displacement
detectors might be mounted between components to detect relative
displacement, e.g., within the reactor vessel. Several transducers
of different types are typically mounted during the pre-operational
testing of a nuclear power plant and then removed. Longer life
sensors of similar type and in similar locations may be used to
provide information in addition to or in lieu of that provided by
existing sensors.
The reactor vessel diagnostics unit 22 may consist of conventional
hardware running software as described below. The hardware in the
reactor vessel diagnostic unit 22 may be a conventional
microcomputer or minicomputer with conventional interfaces to the
signal communication and isolation means 20 and output devices 24.
For example, an IBM PC AT with 640 kilobytes of memory, floppy
disks, removable magnetic media, IEEE, parallel and serial
interfaces, a monitor, a keyboard, a printer and a plotter in a
nonseismic cabinet may be used to evaluate the vibration of the
reactor internals indicated by the neutron activity detected by the
excore and incore neutron detectors 10, 11 and system data
including at least one of pressurizer pressure, reactor coolant
pump motor current, reactor coolant system pressure, reactor
coolant system temperature and reactor coolant system flow rate, to
diagnose the condition of the vibration of the reactor internals
and core structures. In addition, the vibration of the reactor
internals may be evaluated based upon structural data including at
least one of displacement, acceleration and strain of the core
support structures if some of the additional sensors 26, 28 and 30
have been installed.
The software executed by the reactor vessel diagnostic unit 22
preferably includes an expert system which automatically evaluates
vibration of the reactor internals and the system data to establish
whether the data indicates normal or abnormal behavior and, if the
latter, to identify a possible cause of the behavior. As
illustrated in FIG. 2, data supplied via the signal communication
and isolation means 20 is stored as measured values in facts base
32. Also included in the facts base may be constant plant specific
facts input by keyboard or other means. The data stored in the
facts base 32 undergoes data reduction and analysis 34. This data
reduction and analysis preferably includes generating dimensionless
internal vibration indicators and system indicators from the
neutron activity and system data, respectively. These indicators
may be generated by comparing the measured values with baseline
data obtained by sampling data over a period of time from one or
more operating nuclear reactors. Using the standard deviation of
the baseline data for each of the signals received via the signal
communication and isolation means 20, the indicators can then be
assigned a value of, e.g., normal, high, very high, low or very
low, by comparing the difference between the measured values stored
in the facts base 32 with the baseline data stored as part of the
plant specific data in the facts base 32. In other words, if a
measured value exceeds the corresponding baseline data by between
one and two standard deviations, the indicator for that sensor
would be high, and if more than two standard deviations, very high.
Similarly if the indicator for a value is one to two standard
deviations lower than the baseline data for that sensor, then the
indicator would be low and if more than two standard deviations
lower, the indicator would be very low. Otherwise, the indicator
would be normal.
Use of indicators of this sort permits a number of the rules stored
in a rules base 36 to be generic rules. Plants specific rules for
specific types of instrumentation or known operating
characteristics may also be stored in the rules base 36. The facts
stored in the facts base 32 and the rules stored in the rules base
36 may be derived in a known manner from scale model tests, plant
test data, analytical model results, trending analysis results and
other known techniques.
The data reduction and analysis 34 includes both frequency and time
domain analysis and may include applying one or both of auto and
cross power spectral density functions to the reactor internal
vibration and system data. Where structural data from additional
sensors 26, 28 or 30 is available, the auto and cross power
spectral density functions can also be applied to the vibration of
the reactor internals and structural data. Other forms of data
analysis, such as cepstrum analysis of the neutron activity and one
or more of the pressurizer pressure, reactor coolant pump motor
current, reactor coolant system pressure and displacement,
acceleration and strain of the core structures or application of
Duhamel's theorem to the temperature of the reactor coolant system
can also be used.
All of the data analysis techniques described above are known
methods for analyzing data. By applying these or other data
analysis techniques to sample data from operating reactors,
signatures of specific conditions within the reactor vessel can be
identified. In addition, model testing may be done where known
types of deterioration are introduced into the model and the
resulting signatures are recorded. These signatures and their
causes can then be stored as rules in the rules base 36.
For example, the core support structures in a nuclear reactor
include a core barrel support. The reactor internal vibration
obtained from analysis of the neutron activity, e.g., in the manner
described in U.S. Pat. No. 3,860,480, can be converted to the
frequency domain by a fast Fourier transform to identify a
frequency domain statistical property of the neutron activity in
the form of an internal vibration indicator corresponding to the
frequency spectrum of the neutron activity. If the frequency domain
statistical property of the neutron activity exceeds a value
predetermined by analysis of signatures from sample data, then a
diagnosis may be made indicating the possibility of deterioration
of the core barrel support.
As another example, an additional sensor may be installed to detect
displacement of the core barrel. In data reduction and analysis 34,
a time domain statistical property of the reactor coolant
temperature can be compared with a first value determined from
sample data. If the time domain statistical property of the reactor
coolant temperature exceeds the first value and the displacement of
the core barrel indicated by the vibration of the reactor internals
or the additional sensor is in excess of a second value, then a
diagnosis may be made that a hydraulic anomaly, such as vortex
shedding, is likely occurring.
The rules from the rules base 36 and indicators output from the
data reduction analysis 34 are evaluated by an inference engine 38
in the manner described above. The inference engine 38 may be a
conventional expert system shell such as Texas Instruments'
Personal Consultant Plus for the IBM PC. As indicated by the above
examples, the rules can be used to produce a diagnosis of
conditions of the reactor internals and core structures within the
reactor vessel. Additional rules stored in the rules base 36 can
provide recommendations for repair and replacement at the time of
planned outages to avoid unplaned outages. For example, if
deterioration of the core barrel support is diagnosed, the core
barrel support could be identified as needing repair or replacement
during a reloading of the reactor fuel. Thus, a diagnosis and
action recommendations 40 are output from the inference engine
38.
The many features and advantages of the present invention are
apparent from the detailed specification and thus, it is intended
by the appended claims to cover all such features and advantages of
the system which fall within the true spirit and scope of the
invention. Further, since numerous modifications and changes will
readily occur to those skilled in the art, it is not desired to
limit the invention to the exact construction and operation
illustrated and described. Accordingly, all suitable modifications
and equivalents may be resorted to falling within the scope and
spirit of the invention.
* * * * *