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NuclearPlantJournal.com Nuclear Plant Journal, May-June 2014
Integrated
NIMS
By Jin-Ho Park, Korea Atomic Energy
Research Institute (KAERI).
Jin-Ho Park
Jin-Ho Park is the Director & Principal
Researcher at Korea Atomic Energy
Research Institute. He has done research
on Vibration & acoustic analysis,
R&D on structural safety analysis of
mechanical components
in nuclear power plants,
advanced signal processing
for mechanical system’s
condition monitoring &
diagnosis, development
of vibration reduction
techniques for the plant
piping systems, development
of NSSS integrity monitoring
& diagnosis technologies
such as reactor internals
vibration monitoring, loose part
monitoring, acoustic leak monitoring,
and reactor coolant pump vibration
monitoring.
He is a member of the Korean Society
for Noise and Vibration Engineering.
Park has a B.S. and M.S. in Mechanical
Engineering from Pusan National
University, Busan, South Korea. He has
a Ph.D in Mechanical Engineering from
Korea Advanced Institute of Science &
Technology, Daejeon, South Korea.
Overview
The NSSS (Nuclear Steam Supply
System) integrity monitoring system
is the only one to provide the structural
health monitoring for the reactor pressure
boundary components on an on-line basis.
Conventionally, it is comprised of four
independent sub-systems such as IVMS
(Internal Vibration Monitoring System),
LPMS (Loose Part Monitoring System),
ALMS (Acoustic Leakage Monitoring
System), and RCPVMS (Reactor Coolant
Pump Vibration Monitoring System).
The IVMS is designed for early
detection of the degradation of the
preload condition of the reactor internal
structures. The reactor internal structures
consist of core support barrel assembly,
upper guide structure, core
shroud assembly, low support
structure, and hold down
ring. They are subjected to
flow-induced vibration due
to the high pressure reactor
coolant flow. The flow-induced
vibration of the core barrel
assembly may cause the
degradation or loss of the axial
preload at the upper support
flange in the reactor pressure vessel. It
can also result in the loosened or detached
parts inside the reactor vessel. This might
cause significant core damage or coolant
flow blockage in the fuel channel. Thus
IVMS is mainly monitoring the change
of vibratory modal frequencies of the
core barrel assembly to early detect the
degradation of the axial preload by using
ex-core neutron noise signal.
The presence of a loosened or
detached metallic loose part within the
reactor pressure boundary can give rise
to the degradation of the reactor system’s
structural integrity. That is, not only it
may result in the mechanical damage and
fretting wear due to its repeated impact
on the system, but also can cause a partial
flow blockage inside the fuel channel, a
potential for control rod jamming, and the
accumulation of radioactive substances
in the primary system. Thus the primary
purpose of the LPMS is to detect the
presence of a metallic object within the
reactor coolant system using the vibration
sensors installed on the outer surface of
the system. Ultimately, it should give as
useful information as possible to identify
the loose parts such as discrimination
from false detection, localization of the
object and estimation of the structural
effect on the pressure boundary due to its
impact.
The primary purpose of the ALMS is
to monitor coolant leakage in the potential
leak regions such as the reactor vessel,
welded region in pipings, and valves, etc.
The second purpose is to detect initiation
of crack on the surface of the pressure
boundary of the reactor coolant system.
The leak detection is very important
since the leakage could cause a loss of
coolant accident. The ALMS normally
consists of two subsystems. The first one
is for monitoring the opening status of
the PSV (Pressurizer Safety Valves) by
the flow through the valves which had
been mandatorily recommended by U.S.
Nuclear Regulatory Commission Guide
1.45 (Reactor Coolant Pressure Boundary
Leakage Detection Systems). It is called
as PSV system. The second one is for
detecting leaks and cracks in the reactor
coolant system pressure boundary at
specified sensor locations. It is called as
non-PSV system, where AE (Acoustic
Emission) technique has been used to
detect the stress waves caused by the
occurrence of a crack in a solid structure
and the occurrence of fluid leakage.
Theprimary functionof theRCPVMS
is to monitor the shaft displacement and
the rotational speed of a reactor coolant
pump shaft and to monitor the vibration
level of the RCP frame. The RCPVMS is
designed to provide an alarm signal to the
Main Control Room when the vibration
level exceeds the allowable limit. It also
provides diagnostic information to be used
in analyzing the status of the RCP frames
and their shafts, detecting the abnormal
symptom of the shaft crack, and adjusting
the RCP alignment and rotor balancing.
In this system, two types of sensors are
used. One is an accelerometer and the
other is a proximity probe. Typically
three accelerometers are mounted on the
RCP frame to measure the horizontal and
axial vibration levels of the RCP frames.
Three proximity probes are mounted
around the RCP rigid coupling. Two of
them are used to measure orbit (path of
shaft centerline motion) and the vibration
level (displacement) of the RCP shaft.
One proximity probe (keyphasor) is used
to measure the rotating speed and rotating
phase angle of the RCP shaft.
