# Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

Authors: N. Dharmaraju; A. Rama Rao
Journal: Science and Technology of Nuclear Installations
 (2008)

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## Abstract

The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.


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## Body

## 1. Introduction

In the first
phase of the Indian nuclear power programme, the focus was to construct series
of pressurized heavy water reactors (PHWRs) of 235 MWe and 500 MWe ratings in
different part of the country. Presently there are 12 PHWRs of 235 MWe in
operation and additional four units of similar capacity are under advanced
stage of construction. Backed with excellent experience of design, construction,
and operation of 235 MWe type reactors, India has stepped onto indigenous
design of 540 MWe reactors. The first unit of 540 MWe reactors went into
operation successful in September 2005. There are many innovative and
evolutionary design features implemented in almost all the systems so as to
enhance safety in operation, to extend the useful life of the plant, and to
make the operation more users friendly. One such component with new design
features is the coolant channel. It forms an important part of primary heat
transport system. There are 392 coolant channels in the core mounted
horizontally through the calandria and end shield. Each channel carries 13 fuel
bundles in the center, one fuel locator, one shielding plug, and one sealing
plug on the either ends. The fuel bundle is made of 37 cylindrical fuel
elements of 13.22 mm diameter and 495 mm in length. These are held by two end
plates to which the elements are spot-welded. The pressurized heavy water
coolant flows through the coolant channel to carry the nuclear heat to the
steam generators.The coolant
channel and its internals are under constant vibration due to internal coolant
flow and the external moderator circulation. In order to qualify the design of the coolant channel for the flow-induced
vibration excitation in the core, it is important to carry out dynamic
characterization of the coolant channel through vibration studies. Experimental and finite-element modal analyses have been carried out on a representative
coolant channel, its 37 element fuel bundles and the fuel locators under
simulated normal operating condition. This paper highlights the results and its
potential use as diagnostic tool to monitor the health of coolant channel over
its operating life in the core.


## 2. Schematic of Coolant Channel and the Internals

Figure1
shows the schematic of the coolant channel and its internals. The end-to-end
length of the channel is 11.625 m. The outer calandria tube and the internal
pressure tube are both made of zircloy material. The end fittings are made of
stainless steel. To keep the hot pressure tube and the warm calandria tube
separated, there are four garter springs in the annular space. Important
dimensions of the channel are as follows.
(i)
Fuel bundle: 37 element, 496 mm
long, weighing 26 kg/bundle, and 13 bundles/channel.(ii)
Pressure tube: 112 mm OD, 6.425 m long.(iii)
Calandria tube: 132 mm OD, 5.944 m long.(iv)
Fuel locator: 990 mm long and weight 42 Kg (two per channel).(v)
Shielding plug: 990 mm long and weight 56 Kg (two per channel).(vi)
Garter Springs: 4 tight fit on the pressure tube.
On a full-length high temperature
and pressure test set up, vibration testing for modal analysis and vibration
measurement was carried out with rated coolant flow conditions to assess the
severity.For direct measurement on the pressure tube, circular holes were
drilled on the calandria tube using trepanning tool at 6°clock positions. The
positions of these holes were so chosen that they do not foul with the garter
spring position in the annulus but are enough to pick up the response
corresponding to higher modes of the pressure tube. Accelerometers were fixed
on the pressure tube through the holes in the calandria tube and signals were
acquired for online modal and vibration analysis.Figure 1
Coolant channel and its internals.


## 3. Modal Analysis

The pressure
tube, the calandria tube, and the end fittings were instrumented with
accelerometers of sensitivity 100 mvolt/g and excited with portable reaction
type electromagnet shaker. Sweep sine signal in the band of 5 to 100 Hz was
given to the shaker mounted on one end fitting. Figure2 shows a typical force
input spectrum to the end fitting up to 50 Hz. The shape of the spectrum is
typical to the end fitting dictated by its impedance offered to the shaker [1].
Figure 3 shows the frequency response function of the two ends fitting which is
essentially an overhang beam. The first mode natural frequency of the two end
fittings are 10.20 Hz and 15.25 Hz, respectively. Though the two end fittings
are identical, the difference is basically because of the orientation of feeder
pipes connected to two end fittings. One is horizontally connected and the
other is vertically connected.Figure 2
Shaker excitation to the coolant channel.Frequency response function of pressure tube.
(a)(b)(c)Figure3
also shows frequency response function of the pressure tube vibration measured
directly on the pressure tube. The first two modes are at 6.5 Hz and 16.70 Hz.
The pressure tube modes are also seen in the frequency response function of the
end fitting. That is to say that without direct access to pressure tubes in the
reactor, it is still possible to monitor the dynamics of the pressure tube by
monitoring vibration of the end fitting. Figure 4 shows the experimental mode
shape of the pressure tube. The unsymmetrical shape is due to as-installed
difference in the end condition on the two ends of the channel. Such
unsymmetrical shape could be there in the actual reactor channel as well.Experimental mode shapes of coolant channel.
(a)(b)


## 4. Finite-Element Model of Coolant Channel

A finite-element
model of the coolant channel was developed on standard FE package by taking
into account all the loading and boundary conditions. Figure5 shows the FE
Model. The pressure tube, calandria, and end fittings are discretized using
2-node axisymmetric-harmonic structural shell (shell 61) element. The
individual mass of the fuel bundles and the water in the channel is uniformly
distributed on the pressure tube [2]. The mass of fuel locator(s), shielding
plug(s), and sealing plug(s) are lumped at the respective node locations of the
end fitting. The grater springs between the pressure tube and calandria tube
were modeled as rigid links. Modal
analysis has been carried out for the design as well as for other boundary
conditions of the channel as shown in Figures 6 and 7. These conditions are
explained in Table 1. These conditions are selected on the basis of past
experience and in-service inspection in some channels of different operating reactors.
Some of these conditions are realizable in a coolant channel during its
lifetime in the reactor. Others, like shifting of garter springs, missing of
one or more garter springs, and bunching of garter springs, are considered only
for academic exercise (cases 6, 9, and 10). These conditions are not feasible in
the current generation PHWRs under normal circumstances. One more important
condition likely in the channel is irradiation-induced creep. The effect of
creep is to induce sag in the coolant channel. More on this is discussed in
para 6.1. The first three natural frequencies for different conditions of Figures
6 and 7 are listed in Table 2.Table 1
Channel cases considered in the analysis. See Figures6 and 7 for details.
Case 1Channel in design condition and other conditions as per
designCase 2Channel with 3 garter springs and other conditions as
per designCase 3Channel with 2 central garter springs and other
conditions as per designCase 4Channel with 2 outer garter springs and other conditions
as per designCase 5Channel with 1 central garter spring and other
conditions as per designCase 6Channel with NO central garter springs and other
conditions as per designCase 7Channel with only one bearing support (outer) at the
left end-fittingCase 8Channel with only one bearing support (inner) at the
left end-fittingCase 9Channel with NO garter springs and one bearing support
(outer) at the leftCase 10Channel with NO garter springs and one bearing support
(inner) at the leftCase 11Channel with 4 garter springs displaced to center of the
channelCase 12Channel with 2 pairs of garter springs displaced away
from centerTable 2
Natural frequencies of coolant channel for fault scenarios.
CaseFrequency (Hz)1st mode2nd mode3rd modeCase 1 (design)6.4517.6232.66Case 26.4217.0530.05Case 36.3616.6427.12Case 46.4117.5925.65Case 56.2515.6927.11Case 64.9913.5527.30Case 75.5216.6725.82Case 84.0217.5331.92Case 92.6913.7722.73Case 102.6613.5526.43Case 116.2616.1526.74Case 126.4117.5929.71Table 3S. no.ChannelYear of operationYear of contact19951996199720001L-1419841996—8.125 Hz——2M-0719842000—8.030 Hz8.030 Hz8.156 Hz3O-0219841997—8.060 Hz8.125 Hz—4P-1319841996—8.156 Hz8.156 Hz—5D-151986—8.06 Hz———6N-101986—8.03 Hz———Figure 5
Finite element model of the coolant channel.Figure 6
Coolant channel different garter spring fault scenario. G1 to G4 garter springs. Bearings shown in green.Figure 7
Coolant channel with different bearing faults and garter spring fault scenario.


### 4.1. Channel Dynamics in Design Condition

The first two natural frequencies of the
coolant channel model are 6.45 Hz and 17.62 Hz. These frequencies closely match
with the experimental values shown in Figure4 under simulated loading and
boundary conditions. As the experimental channel is a full-scale setup, the
close match of FE results with experimental results also ensures that the FE
results are directly applicable to the reactor. Having tuned the FE model, it
can be used for dynamic study under simulating loading condition such as fluid
structure interaction, seismic analysis, shock response.The important boundary conditions of direct
relevance and that which can change the dynamics of the channel significantly
are cases number 7 and 8, which pertain to effectiveness of bearings supports
to the end fittings. It is quite likely that one or more bearing (out of four)
may become partially ineffective during the lifetime of the channel or when the
irradiation-induced sag in the channel is large. The loss of bearing support to
the coolant channel reduces the channel frequencies significantly. Such a
change is measurable and hence detectable for monitoring the condition of coolant channels.


## 4.1. Channel Dynamics in Design Condition

The first two natural frequencies of the
coolant channel model are 6.45 Hz and 17.62 Hz. These frequencies closely match
with the experimental values shown in Figure4 under simulated loading and
boundary conditions. As the experimental channel is a full-scale setup, the
close match of FE results with experimental results also ensures that the FE
results are directly applicable to the reactor. Having tuned the FE model, it
can be used for dynamic study under simulating loading condition such as fluid
structure interaction, seismic analysis, shock response.The important boundary conditions of direct
relevance and that which can change the dynamics of the channel significantly
are cases number 7 and 8, which pertain to effectiveness of bearings supports
to the end fittings. It is quite likely that one or more bearing (out of four)
may become partially ineffective during the lifetime of the channel or when the
irradiation-induced sag in the channel is large. The loss of bearing support to
the coolant channel reduces the channel frequencies significantly. Such a
change is measurable and hence detectable for monitoring the condition of coolant channels.


## 5. Response of the Channel due to Coolant Flow

The
channel response was measured during the rated coolant flow of about 30 Kg/sec
at room temperature. Figure8 shows the response of the pressure tube due to
the flow excitation. The first two pressure tube modes at 6.25 Hz and 16.7 Hz,
respectively, can be clearly seen. It has been seen that the amplitudes of the
two modes decrease for lower than rated flow due to lower flow excitation [3].
The amplitude of pressure tube vibration during normal rate of coolant flow
then was 6.3 microns.
Figure 9 shows the response of the two end fittings. The dominant peaks in the
two spectra correspond to the first mode natural frequencies of the two end
fittings. These modes were identified as two end-fitting modes as explained in
para 3.0 and shown in Figure 3. The first mode of the pressure tube can also be seen
in the response spectrum of one of the end fitting at 6.25 Hz with small but distinct amplitude.
The amplitude ratio of 6.25 Hz peak at center of the pressure and on the end
fitting is 10 : 1 [4]. With modern sensors and signal analysis capabilities, it
is possible to identify the first mode of the pressure tube on the end
fitting.Figure 8
Flow excitation spectrum of pressure tube.Flow excitation spectrum of end fittings.
(a)(b)


## 6. Channel Response Characteristics as a Tool for Diagnostics

Having been able to measure the
pressure tube vibration from the end fitting, this possibility opens the potential
of diagnosing the health of the pressure tube during plant operation. In PHWR
type of reactor, the fuel loading is an online process carried out by the
special purpose machine, which latches onto the channel to replace the fuel
bundle. The same machine can be made use of for measuring the end fitting
response. If it is not found suitable, a separate campaign may be undertaken.
The only requirement would be to operate primary coolant pumps. It has been
observed that the shut down flow is not sufficient to excite the channel mode
to the extent that the channel frequency can be detected on the end fittings.As can be
seen from the results of FE model of the coolant channel, any internal changes
from as-installed/design condition of the channel results in change in its
modal frequencies. The first mode is seen to be more sensitive to changes in
the channel-bearing conditions and number of effective garter springs in the
annulus. Diagnostics developed on the basis of channel vibration to detect changed
garter spring position in earlier generation PHWR reactor is explained in next
section.


### 6.1. Nonintrusive Diagnostic of Coolant Channels in Earlier
Generation PHWR

In the
earlier generation Indian PHWRs, the concept coolant channels are similar to
the new 540 MWe PHWR channels except that the sizes are different. Accordingly,
the modal frequencies of the channel are different. There were two loose garter
springs in the annulus of the coolant channel instead of four tight-fit type as
in 540 MWe channel to keep the hot pressure tube and the cold calandria tube
physically separated. The fuel used in these reactors was 19 element bundles
weighing about 16 Kg, which are similar to CANDU’s 19 element fuel bundles.During the
commissioning activity, when the channels are subject to be hot conditioning,
there is no fuel in the channel. In such a condition, the loose garter springs
are free to slide on the pressure tube. For hot conditioning, hot fluid
flows in the pressure
tubes and hence induces vibration in the tube. Due to flow-induced vibrations
in the tube, the garter springs got shifted in number of channels. This
displacement resulted in lowering the design gap of 8.0 mm between the two
tubes when the fuel is loaded. This and irradiation-induced creep in the tube
led to premature contact between the tubes in loaded condition. Such a contact
is considered to be hazardous for the safety of the reactor.To identify
the contacting channels in the core, a nonintrusive vibration technique was developed
based on the change in natural frequency of the channel. In this technique, one
end of the channel (end-fitting) is excited by a portable shaker and response
is measured from the other end of the channel. The excitation was tuned to
excite only the first mode of the channel. With this technique, the entire core
was screened to detect change in the first natural frequency. As low-level
excitation is given to the channel, the neighboring channel does not get
affected nor do they influence the channel under excitation. Such screening
campaign was carried out on more than 2000 channels, repeating two to three
times on the same channel over the period of several full power years. It was
found that in the channels with high creep, the first natural frequency of the
channel increased to a higher value. The table below shows the trend of the
natural frequencies with high and low creep channels.The first
mode natural frequency of the coolant channel of 235 MWe PHWR is 7.95 Hz as
against 6.5 Hz for the 540 MWe channels. Due to irradiation-enhanced creep and
due to displaced garter springs, some of the channels had sagged (elastic and
plastic) by more than 14 mm in the center. The deformation due to creep had set
the channels into a permanently sagged condition. Such channels showed an
increase in the natural frequency. The pace of increase was found to be very
slow in a healthy channel whereas
in a highly sagged channels, the increase was found to be relatively fast.Some of the
typical channel frequencies and its changes over the period of time are shown
in the table below. All the channels shown in the table below were removed from
the core in the respective years for postirradiation inspection of suspected
contact based on analytical estimation of time of contact. In these, some are
high flux and some are low flux channels. Except for channel D15 and N10 other
four channels were found contacting. It may be seen from the table that all the
four contacting channels had first-mode frequency which was equal to or more than 8.125 Hz , which
was an increase of 0.175 Hz from normal.
Since the smallest measurable frequency by the analysis system was 0.03 Hz, it
was possible to trend the change in the frequency accurately. Hence, the data
bank of the channel natural frequency was reliably used as a diagnostic tool to
identify healthiness of a channel.


## 6.1. Nonintrusive Diagnostic of Coolant Channels in Earlier
Generation PHWR

In the
earlier generation Indian PHWRs, the concept coolant channels are similar to
the new 540 MWe PHWR channels except that the sizes are different. Accordingly,
the modal frequencies of the channel are different. There were two loose garter
springs in the annulus of the coolant channel instead of four tight-fit type as
in 540 MWe channel to keep the hot pressure tube and the cold calandria tube
physically separated. The fuel used in these reactors was 19 element bundles
weighing about 16 Kg, which are similar to CANDU’s 19 element fuel bundles.During the
commissioning activity, when the channels are subject to be hot conditioning,
there is no fuel in the channel. In such a condition, the loose garter springs
are free to slide on the pressure tube. For hot conditioning, hot fluid
flows in the pressure
tubes and hence induces vibration in the tube. Due to flow-induced vibrations
in the tube, the garter springs got shifted in number of channels. This
displacement resulted in lowering the design gap of 8.0 mm between the two
tubes when the fuel is loaded. This and irradiation-induced creep in the tube
led to premature contact between the tubes in loaded condition. Such a contact
is considered to be hazardous for the safety of the reactor.To identify
the contacting channels in the core, a nonintrusive vibration technique was developed
based on the change in natural frequency of the channel. In this technique, one
end of the channel (end-fitting) is excited by a portable shaker and response
is measured from the other end of the channel. The excitation was tuned to
excite only the first mode of the channel. With this technique, the entire core
was screened to detect change in the first natural frequency. As low-level
excitation is given to the channel, the neighboring channel does not get
affected nor do they influence the channel under excitation. Such screening
campaign was carried out on more than 2000 channels, repeating two to three
times on the same channel over the period of several full power years. It was
found that in the channels with high creep, the first natural frequency of the
channel increased to a higher value. The table below shows the trend of the
natural frequencies with high and low creep channels.The first
mode natural frequency of the coolant channel of 235 MWe PHWR is 7.95 Hz as
against 6.5 Hz for the 540 MWe channels. Due to irradiation-enhanced creep and
due to displaced garter springs, some of the channels had sagged (elastic and
plastic) by more than 14 mm in the center. The deformation due to creep had set
the channels into a permanently sagged condition. Such channels showed an
increase in the natural frequency. The pace of increase was found to be very
slow in a healthy channel whereas
in a highly sagged channels, the increase was found to be relatively fast.Some of the
typical channel frequencies and its changes over the period of time are shown
in the table below. All the channels shown in the table below were removed from
the core in the respective years for postirradiation inspection of suspected
contact based on analytical estimation of time of contact. In these, some are
high flux and some are low flux channels. Except for channel D15 and N10 other
four channels were found contacting. It may be seen from the table that all the
four contacting channels had first-mode frequency which was equal to or more than 8.125 Hz , which
was an increase of 0.175 Hz from normal.
Since the smallest measurable frequency by the analysis system was 0.03 Hz, it
was possible to trend the change in the frequency accurately. Hence, the data
bank of the channel natural frequency was reliably used as a diagnostic tool to
identify healthiness of a channel.


## 7. Fuel Bundle Vibration

The fuel
bundle in the 540 MWe reactors is made of 37 elements, which is similar in construction
to standard CANDU fuel. In Indian PHWR channels, there are some changes with
respect to larger CANDU reactor wherein there are 13 fuel bundles per channel
instead of 12 and there is fuel locator on the upstream and downstream of the
13 bundles, which is free in the channel and so is in contact with the first
and the last fuel bundle. The need for a fuel locator has come as a requirement
for the fueling machine design.In view of
the new in-core components in the channel, a study was carried out to
investigate its effect on fuel vibration performance. Such a study was carried
out in a specially erected low-temperature setup to measure fuel and fuel
locator vibration by optical vibration transducer through a Perspex window cut
in the setup. All the fuel boundary conditions were simulated and the dynamic
head of the coolant flow was maintained in the setup. Flow turbulence being the
main source of excitation to the in-core components, the fuel and fuel locator
vibrations were measured at different flow rate through the channel.Figure10 shows the picture of
optical probe measuring the fuel vibration. The figure also shows the
three-vibration spectrum of the first bundle measured at the front, middle, and
at the rear of the bundle. The dominant (marked) peak in the spectrum
corresponds to the first bending mode of the fuel element, which is at 36 Hz.
The bending mode of the element is excited at about 70% of the rated flow. It
was observed that the frequency band of flow excitation reduces with increase
in flow rate [5]. At 100% flow, the excitation remains confined up to 30 Hz.
The rocking mode [5] of the bundle at 2.62 Hz can be seen in the three spectra
shown in Figure 10. The amplitude of
vibration of the element at 36 Hz was about 10 micron.Figure 10
Experimental results of 37 element bundle vibration under
flow excitation.


## 8. Conclusions

The coolant
channel is the heart of PHWR type of reactors. In PHWR, normal operating life
of a channel is 20–25 years. Besides
the wear and tear in the channel due to continuous operation, they are also
subjected to irradiation-induced damages to the material of the channel and
associated substructures. It is hence desirable to monitor the changes
happening in the channel from the reactor safety point of view. The easiest
method is to check changes in the dynamics of the channel. The experimental and modeling study on the
coolant channel has been carried with an aim to understand the dynamics of the
channel. The measured natural frequencies identified in the modal analysis and
from the response analysis of the channel during flow excitation can be used
for monitoring the health of the channel.Changes in
the support conditions (garter springs), large sagging deformation, premature
contact, and other mechanistic changes also influence the dynamics of the
channel. A number of such changes/deviations have been observed in Indian PHWR
channel over its operating life. These changes limit the safe operating life of
the channel to less than 15 years. The study carried out in this paper aims to
demonstrate the possibility of detecting the changes that happen in the
channel. As an example, detecting contact between hot pressure tube and cold
calandria has been illustrated based on actual measurement and inspection.The fuel
bundles in the channel are also subjected to flow-induced vibration. It has
been demonstrated in the paper that under simulated test conditions, it is
possible to detect fuel vibration. Such an investigation is especially useful
during the design phase of the fuel. There are number of reported fuel failures
in operating nuclear reactors due to flow-induced vibration. Detection of the
bending mode of the fuel element is perhaps reported for the first time through
this paper. Such fuel vibration overextended period of time may result in fret-related
damages in the fuel and in the channel that support the fuel.



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*Source: 764301-2008-04-01.xml*