Information Notice 1999-14, Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrick
UNITED STATES NUCLEAR REGULATORY
COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001 May 5, 1999 NRC INFORMATION
NOTICE 99-14: UNANTICIPATED
REACTOR WATER DRAINDOWN AT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONE UNIT 2, AND FITZPATRICK
Addressees
All holders of licenses for nuclear power, test, and research reactors.
Purpose
The U.S. Nuclear Regulatory
Commission (NRC) is issuing this information
notice to alert addressees
to the potential
for personnel
errors during infrequently
performed
evolutions
that result in, or contribute
to, events such as the inadvertent
draining of water from the reactor vessel during shutdown operations.
It is expected that recipients
will review the information
for applicability
to their facilities
and consider actions, as appropriate, to prevent a similar occurrence.
However, suggestions
contained
in this information
notice are not NRC requirements;
therefore, no specific action or written response to this notice is required.DescriDtion
of Circumstances
Quad Cities Unit 2 On February 24, 1999, Quad Cities Unit 2 was in cold shutdown with reactor water temperature
at 131 'F and reactor water level at 80 inches indicated
level (normal level during operations
is 30 inches indicated
or 173 inches above the top of active fuel [TAF]). Core cooling was being maintained
in a band of 120 'F to 170 OF by the OA" loop of the shutdown cooling mode of the residual heat removal (RHR) system after being switched from the nB" loop at 12:32 a.m.During the switch over the licensee inadvertently
failed to close the OA RHR minimum flow valve as required by the procedure.
Sometime later operators
noted a decreasing
reactor water level and at about 1:02 a.m. secured the *2A RHR pump and isolated shutdown cooling. At 1:55 a.m. operators
restored the *2A' loop of shutdown cooling to the proper lineup and started the *2A RHR pump. Water level had decreased
to a minimum of about 45 inches indicated, and reactor water temperature
had risen to a maximum of about 163 OF. Forced circulation
of reactor vessel water using a reactor recirculation
pump remained in effect throughout
the event.On the basis of post event reviews, It appears that the minimum flow valve in the OA loop was left open because the nuclear station operator failed to ensure that the tasks were performed
in the sequence specified
in the operating
procedures.
The nuclear station operator who was (7008 PD(L H ort<<4qj-Oiif
qqos(J5 C7Ffcj ANW\\b
IN 99-14 May 5, 1999 directing
the evolution
from the control room gave the non-licensed
operator permission
to de-energize the breaker for the WA RHR minimum flow valve operator before the valve was taken to the required closed position.
De-energizing
the breaker also removed power to the valve position indicator
lights in the control room. Thus, when the nuclear station operator tried to verify that the valve was closed, there was no position indication
in the control room to make that verification.
The nuclear station operator made the incorrect
assumption
that the valve was already closed and moved to the next step in the procedure.
This failure to close the WAX RHR minimum flow valve opened a drain path from the reactor to the suppression
pool. To further complicate
the event, the operating
crew did not recognize
that there was any problem until approximately
10 minutes had passed and the water level had decreased
about 13 inches because of a misinterpretation
of causes of the level decrease.
After detecting
the decrease, the operating
crew was slow to react, which allowed the level to decrease another 20 inches before the operators
isolated shutdown cooling which terminated
the draindown.
The licensee estimated
that a total of 6000 to 7000 gallons was drained from the reactor to the suppression
pool.Operations
staff practices
including
poor communications, poor activity briefings
for high-risk activities, lack of effective
pre-shift
briefings, inadequate
supervision
of important
control room activities, inadequate
monitoring
of control room panels, and slow event response may have contributed
to the event. Although the unintended
loss of inventory
to the suppression
pool highlighted
significant
weaknesses
in plant operations, the safety significance
was minimized
by two features.
First, a reactor recirculation
pump remained in service throughout
the event which served to distribute
decay heat. Second, an automatic
isolation
of shutdown cooling would have occurred at 8 inches indicated
level which would have stopped the draining event.An indicated
water level of 8 inches corresponds
to approximately
151 inches of water level above the TAF in the reactor core.Arkansas Nuclear One Unit 2 On February 2, 1999, at Arkansas Nuclear One Unit 2, the operators
were draining the refueling
canal in preparation
for installing
the reactor vessel head. Refueling
was complete and steam generator
nozzle dams were installed.
The operators
were using the two low pressure safety injection (LPSI) pumps to drain the canal to the refueling
water storage tank;one pump also served as the shutdown cooling pump. The rate of draindown
was approximately
3.3 Inches per minute. When the water level reached 105 inches, the reactor operator noted that level started to lower rapidly. Operators
stopped one of the LPSI pumps and instructed
a local operator to close the isolation
valve to the refueling
water tank. This manually operated valve required 55 turns of the handwheel
to fully close. Within approximately
1.5 minutes, the reactor vessel level had dropped below the 65 inch level (where reduced inventory
begins) and continued
down to 56 inches before the valve could be fully closed. (Reference
zero on these level instruments
is the bottom of the hot leg, with mid-loop being defined at approximately
24 inches.) The average rate of level decrease between 105 IN 99-14 May 5, 1999 inches and 56 inches was approximately
33 inches per minute. At its lowest level, 56 inches indicated, there were still 93 inches of water above the TAF. Using the high pressure safety injection (HPSI) pump the operators
brought the level back up to 90 inches. The plant was in reduced inventory
operations (below 65 inches) for approximately
7 minutes. During the event the level remained well above the point where LPSI pump cavitation
would be expected.
The licensee concluded
that the safety significance
of the event was minimal because multiple sources of makeup water were available, redundant
mitigation
equipment
was available, and the operators
were quick to recognize
and respond to the event.On the basis of post event reviews, it was determined
that the procedure
used for draining down the refueling
canal was inadequate
in that it incorrectly
stated that the draindown
should be secured at the 90-inch level. The procedure
should have directed that the rate of draining be secured at the 106-inch level so that appropriate
precautions
could be taken before resuming the draindown.
These precautions
should have Included reminders
to the operating crew that below the 106-inch level the level will drop much more quickly due to the transition
of pumping from a large volume in the refueling
canal to a small volume In the reactor vessel.Therefore, in order to maintain control of the water level, the draindown
rate should be decreased
and an operator should be stationed
to directly monitor the level.Additional
factors that contributed
to this event include: the operators
received little specific training on this evolution;
the crew was inexperienced
in performing
this task; the task should have been classified
as an infrequent
task requiring
a more thorough briefing;
and, operators failed to station an operator in a position where he could directly monitor the water level in the refueling
canal. Instead they monitored
it remotely using a video camera that did not provide a clear picture of the water level.FitzPatrick
On December 2, 1998, at the James A. FitzPatrick
Nuclear Power Plant, the operators
were in the process of reassembling
the reactor following
refueling.
Operators
were controlling
the reactor vessel water level at 357 inches above TAF by adjusting
the water discharge
rate to compensate
for the constant input from the control rod drive cooling water system. While in this condition, the licensees
risk analysis requires that reactor vessel water level be monitored
using two independent
level indicators.
To meet this requirement, the licensee designated
a wide range indicator
which provided Indication
up to the top of the reactor vessel and an RHR interlock
level indicator
which provided indication
in the range from -150 inches to +200 Inches as the instruments
to be used during this evaluation.
In order for the wide-range
level Indicator
to remain available
with the reactor head removed, a temporary
standpipe
and fill funnel were used to replace a portion of the reference
leg. At the time of the event, the licensee was in the process of removing this temporary
standpipe
and reinstalling
the original reference
leg components.
As the water drained from the standpipe, it caused the wide-range
level indicator
to erroneously
show an increasing
water level. For a period of approximately
one hour the operators
in the control room, unaware that the ongoing maintenance
would cause an error in the indicated
water level, compensated
for the apparent increasing
level by increasing
the discharge
rate. This action had the effect of reducing the
IN 99-14 May 5, 1999 actual water level from 357 inches to 255 inches. During the same time period, the operators were also in the process of filling and venting the reactor feedwater
piping, which could have affected the reactor water level. Once the normal reference
leg piping had been reinstalled
and the reference
leg began to refill, the indicated
level decreased
from 357 inches to the actual level of 255 inches. The second level instrument, which does not come on-scale until the level goes below 200 inches, remained off-scale
high.When operators
discovered
the level discrepancy, they used a temporary
pressure gauge connected
to the reactor vessel low-point
tap to confirm the actual water level. After confirming
the accuracy of the wide-range
indicator, they restored the reactor vessel water level to 357 inches. The 100-inch error represented
approximately
14,000 gallons of water. The licensee determined
that the safety significance
of this event was low since the reactor was in cold shutdown with low decay heat and the reactor water level remained well above the TAF. In addition, the drain-down
would have been limited by an automatic
Isolation
of the draindown path, which would have occurred prior to vessel level reaching 177 Inches above the TAF.The licensee's
post event review identified:
weaknesses
in the operator's
knowledge
of the reactor assembly process; lack of explicit detail in the reactor assembly procedure;
and, weaknesses
in the plant risk assessment
process. Contrary to the assumption
that two designated
reactor water level indicators
were available, only one indicator, the wide-range
instrument, was available
in the range above 200 inches. When the reference
leg on the wide-range instrument
was disassembled
and drained, the one usable indicator
was rendered unavailable.
The second instrument
was pegged off-scale
high and remained that way throughout
the event because the level never dropped below 200 inches. A post event review by the licensee indicated
that other reactor water level instruments, remained operable during the event but, apparently
the operators
did not rely on these other instruments
or notice the discrepancy
between them and the wide range Indicator.
Proposed corrective
actions included procedural
enhancements
to ensure that reactor level instrumentation
credited by the outage risk assessment
remains available
during reactor disassembly
and reassembly.
Discussion
Personnel
errors appear to have caused, or contributed
to, these three inadvertent
reactor vessel draindown
events. The likelihood
of personnel
errors is dependent
upon the operators knowledge
of the task gained through previous experience
and training.
It is also dependent upon the quality of the procedures
used to perform the task, the level of supervision, the adequacy of pre-job briefings, fatigue, and distractions
resulting
from multiple tasks. In each of the events, the plant staff made errors during a seldom-performed
evolution.
Because it was a seldom-performed
evolution, more training, better pre-job briefings, closer supervision, and procedures
that contain more details than those for frequently
performed
activities
might have prevented
these events.
IN 99-14 May 5, 1999 This information
notice requires no specific action or written response.
If you have any questions
about the information
in this notice, please contact the technical
contact listed below, the appropriate
regional office, or the appropriate
Office of Nuclear Reactor Regulation (NRR)project manager.Ledyard B. Marsh, Chief Events Assessment, Generic Communications
And Non-Power
Reactors Branch Division of Regulatory
Improvement
Programs Office of Nuclear Reactor Regulation
Technical
contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdDRenrc.aov
REFERENCES:
NRC Integrated
Inspection
Report No. 50-333/98-08, issued February 10, 1999 (Accession
No.9902170348)
for the James A. FitzPatrick
Nuclear Power Plant for the period November 22, 1998, through January 10, 1999.Attachment:
List of Recently Issued NRC Information
Notices
~~ Attachment
1 IN 99-14 May 5, 1999 Page 1 of I LIST OF RECENTLY ISSUED NRC INFORMATION
NOTICES Information
Date of Notice No. Subject Issuance Issued to 99-13 Insiahts from NRR Inspections
4129199 All holders of operatina
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Circuit Breaker Maintenance
Programs for nuclear power reactors 99-12 Year 2000 Computer Systems Readiness
Audits Incidents
Involving
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4/28/99 4/23/99 All holders of operating
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removed from the reactor 99-10 99-09 Degradation
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When 3/24/99 Manually Editing Treatment
Data on The Nucletron
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4/1/99 All holders of operating
licenses for nuclear power reactors All medical licensees
authorized
to conduct high-dose-rate (HDR)remote after loading brachytherapy
treatments
All holders of operating
licensees for nuclear power reactors and licensees
authorized
to possess or use formula quantities
of strategic
special nuclear material 99-08 OL = Operating
License CP = Construction
Permit
IN 99-xx April xx, 1999 Page 5of 5 This information
notice requires no specific action or written response.
If you have any questions
about the information
in this notice, please contact the technical
contact listed below, the appropriate
regional office, or the appropriate
office of Nuclear Reactor Regulation (NRR)Project Manager.Ledyard B. Marsh, Chief Events Assessment, Generic Communications
And Non-Power
Reactors Branch Division of Regulatory
Improvement
Programs Office of Nuclear Reactor Regulation
Technical
contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdRDanrc.aov
REFERENCES:
NRC Integrated
Inspection
Report No. 50-333198-08, issued February 10, 1999 (Accession
No.9902170348)
for the James A. FitzPatrick
Nuclear Power Plant for the period November 22, 1998, through January 10, 1999.Attachments:
1. List of Recently Issued NMSS Information
Notices 2. List of Recently Issued NRC Information
Notices DOCUMENT NAME: G:ICDPDRAININ\DRAIN.0B.WPD
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DATE V /0199 [3 /1/99 4 /4I9 1' /0g99 F .V. ...OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP
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IN 99-14 May 5, 1999 This information
notice requires no specific action or written response.
If you have any questions
about the information
in this notice, please contact the technical
contact listed below, the appropriate
regional office, or the appropriate
Office of Nuclear Reactor Regulation (NRR)project manager.[arig sjid by]Ledyard B. Marsh, Chief Events Assessment, Generic Communications
And Non-Power
Reactors Branch Division of Regulatory
Improvement
Programs Office of Nuclear Reactor Regulation
Technical
contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdr)ODnrc.gov
REFERENCES:
NRC Integrated
Inspection
Report No. 50-333/98-08, issued February 10, 1999 (Accession
No.9902170348)
for the James A. FitzPatrick
Nuclear Power Plant for the period November 22, 1998, through January 10, 1999.Attachment:
List of Recently Issued NRC Information
Notices DOCUMENT NAME: S:XDRPMSEC\99-14.IN
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