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
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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
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 licenses
of Low-and Medium-Voltage
Circuit Breaker Maintenance
Programs
for nuclear power reactors
99-12
Year 2000 Computer Systems
Readiness Audits
Incidents Involving the Use of
Radioactive Iodine-131
4/28/99
4/23/99
All holders of operating licenses
or construction permits for nuclear
power plants
All medical use licensees
99-11
97-15, Sup 1 Reporting of Errors and
4/16/99 Changes in Large-Break/Small- Break Loss-of-Coolant Evaluation
Models of Fuel Vendors and
Compliance with 10 CFR 50.46(a)(3)
All holders of operating licenses
for nuclear power reactors, except
those who have permanently
cease operations and have
certified that fuel has been
permanently removed from the
reactor
99-10
99-09 Degradation of Prestressing
4/13/99
Tendon Systems in Prestressed
Concrete Containments
Problems Encountered When
3/24/99
Manually Editing Treatment Data
on The Nucletron Microselectron-HDR
(New) Model 105.999 Urine Specimen Adulteration
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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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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