Information Notice No. 86-101:Loss of Decay Heat Removal due to Loss of Fluid Levels in Reactor Coolant System
SSINS No.: 6835
IN 86-101
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, DC 20555
December 12, 1986
Information Notice No. 86-101: LOSS OF DECAY HEAT REMOVAL DUE TO LOSS OF
FLUID LEVELS IN REACTOR COOLANT SYSTEM
Addressees:
All holders of an operating license or a construction permit for
pressurized-water reactor (PWR) facilities.
Purpose:
This notice is intended to advise licensees of continuing problems during
PWR outages with procedures and instrumentation for control of water level
in reactor vessels when reactor coolant systems (RCSs) are partially drained
for maintenance. These problems have resulted in temporary loss of decay
heat removal.
It is expected that recipients will review this information for
applicability to their reactor facilities and consider actions, if
appropriate, to preclude occurrence of similar problems. Suggestions
contained in this notice do not constitute NRC requirements; therefore, no
specific action or written response is required.
Description of Circumstances:
A typical PWR has a decay heat removal system with two redundant trains.
Generally, both trains take suction from the same RCS hot leg, and the
connecting piping is attached to either the bottom or a lower quadrant of
the hot leg. During certain maintenance activities, the water level in the
reactor vessel must be lowered below the tops of the nozzles which connect
the hot legs to the reactor vessel. Lowering the level too far can cause
vortexing in the hot leg at the suction nozzle for the decay heat removal
system, air entrainment in the water flowing to the operating decay heat
removal pump, and air binding of the pump. If the other pump is started, it
too is likely to become air bound. Consequently, all decay heat removal is
lost until the water level in the reactor vessel and hence in the hot leg
piping is raised and the decay heat removal pumps are vented and restarted.
During outages in the last year and half, decay heat removal pumps at
several PWRs lost suction because of vortexing. Four of these events are
described in Attachment 1 to this information notice. Deficiencies which
contributed to the events include: (1) lack of operator knowledge about the
correlation between water level and pump speed at the onset of vortexing and
air entrainment
8612090402
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IN 86-101
December 12, 1986
Page 2 of 2
(2) operating procedures that did not adequately consider vortexing and air
entrainment
(3) reactor vessel water level instrumentation which was erratic or inac-
curate, did not have adequate range, was not checked adequately before
use, or was not monitored as frequently as necessary during use
During one of these events, local boiling of reactor coolant and some
release of radioactive contamination to containment did occur.
Discussion:
In the aggregate, licensees involved in the events described in Attachment 1
have taken certain actions. These actions include additional operator
training, improvement of instrumentation for monitoring water level in the
reactor when the level has been lowered for maintenance, addressing in
operating procedures the relationship between water level and flow rate for
the onset of vortexing and air entrainment, and requiring in operating
procedures that the performance of water level instrumentation be checked
before water level is lowered.
The nuclear industry has been previously made aware of this problem. IE In-
formation Notice 81-09 described an event that occurred at Beaver Valley
Unit in March 1981. Further, the Nuclear Safety Analysis Center operated by
the Electric Power Research Institute published NSAC-52 in January 1983.
This report provides information on 12 PWR events which occurred from 1977
through 1981 and which resulted in the loss of capability to remove decay
heat because of reduction of water inventory in the RCS. Case Study Report
AEOD/C503 issued in December 1985 by NRC's Office of Analysis and Evaluation
of Operating Data presents similar information from 1976 through 1984. That
case study indicates that there were 32 events during that period including
6 in 1984. Although these reports are available to the industry, significant
events continue to occur.
This notice requires no specific action or written response. If you have any
questions regarding this matter, please contact the Regional Administrator
of the appropriate regional office or this office.
Edward L. Jordan, Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: Roger W. Woodruff, IE
(301) 492-7205
Attachments:
1. Loss of RHR events at PWRs
2. List of Recently Issued IE Information Notices
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Attachment 1
IN 86-101
December 12, 1986
Page 1 of 4
LOSS OF RHR EVENTS AT PWRs
San Onofre 2
On March 19, 1986, San Onofre 2, a Combustion Engineering designed reactor,
was in cold shutdown and preparations were made to partially drain the RCS
and perform maintenance in a steam generator channel head. Before initial
draining the reactor vessel to the midpoint of the RCS hot and cold legs,
wide- and narrow-range RCS level instruments were put in service by
installing their temporary connections and calibrating them. Because their
readings oscillated when a portable RCS eductor for control of airborne
radioactive contamination was operated tygon tubing was installed
temporarily to provide a sight gauge for monitoring a er level. Thus, three
devices were available for monitoring water level in the system.
On March 26, the water level in the reactor vessel was below the vessel
flange, the RCS was vented to the containment atmosphere via incore detecter
nozzles in the vessel head, a low-pressure safety injection (LPSI) pump was
running to provide decay heat removal via the shutdown cooling system
(SDCS), and a temporary dam was installed in the cold leg nozzle of the
steam generator to facilitate maintenance which was to be performed on it.
To permit repair of,the nozzle dam which had been leaking, the water level
in the reactor vessel was being lowered to 17.5 inches above the bottom of
the 42-inch diameter hot legs.
One of the hot legs supplies water to the inlet side of the SDCS. The nozzle
for the connecting pipe to the SDCS is located on the bottom of that hot
leg. While the water level was being lowered, a vortex formed on the suction
side of LPSI Pump 16. The vortex entrained air causing the pump to become
air bound, loss of SDCS flow, and thus loss of decay heat removal. To avoid
damage to the pump, it was secured. The redundant pump, LPSI Pump 15, was
started, and it too became air bound and was secured. To again establish
flow through the SDCS, the system was vented, and the water level in the
reactor vessel was raised. Seventy minutes after the first indication of
vortexing, decay heat removal was #gain established when LPSI Pump 16 was
returned to service. During the time that decay heat removal was lost, the
hot leg temperature increased from 114 F to 210 F, and local boiling
occurred in the reactor core. Steam and 2 curies of radionuclides were
released to containment.
The wide- and narrow-range level instruments are connected to taps on the
RCS hot leg drain line and on the pressurizer. Instrument zero for the
narrow-range instrument is at the level of the bottom of the hot leg, and
its range is from zero to +42 inches, i.e., the top of the hot leg.
Instrument zero for the wide-range instrument is at the reactor vessel
flange, and its range is from -120 inches (or 19.5 inches below the bottom
of the inside surface of the hot leg) to +300 inches. The operators distrust
these two instruments because their readings oscillate when the RCS eductor
is operating and because low points in flexible tubing at the upper pressure
tap collect condensate.
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Attachment 1
IN 86-101
December 12, 1986
Page 2 of 4
The RCS eductor is a portable device which is temporarily installed by
maintenance personnel when the RCS is opened for repair work. The eductor
takes suction on the air space above the reactor coolant surface and
discharges to the containment purge system. Its function is to minimize the
exposure of maintenance personnel to airborne radioactive contamination.
While installing and filling the tygon tubing, an air bubble was
inadvertently trapped in the tubing causing it to read high by 10.5 inches.
Further, the reference scale for the tubing was displaced by 2.5 inches in
the upward direction causing a total error of 13 inches on the high side.
The operators were relying on this device while reactor water level was
being lowered. The licensee intended to lower the level to 17.5 inches above
the SDCS nozzle; however, the level was actually being lowered to 4.5 inches
above the SDCS nozzle. After the level reached 9.5 inches, vortexing
started. Although, the operator did not have confidence in the narrow range
instrument, its reading was approximately correct at that time.
The operator did not have at hand a formal correlation of the potential for
vortexing as a function of water level and SDCS flow rate. Lack of knowledge
about the performance of the system at low water levels and unreliable
instrumentation for monitoring water level were the principal causes of this
event.
Zion 2
On December 10, 1985, Zion 2, a Westinghouse designed reactor, was in cold
shutdown with the water level in the reactor vessel below the flange, the
RCS vented to atmosphere, a residual heat removal (RHR) pump running to
provide decay heat removal, and a charging pump running to provide makeup to
the RCS. The water level in the reactor vessel had been lowered to
facilitate repair of an RHR valve. A recorder in the control room.was
connected to the refueling water level transmitter and was being used to
monitor the water level in the reactor vessel.
Between December 10 and 14, enough additional water was inadvertently
removed or lost from the RCS to lower the water level in the vessel far
enough to cause vortexing and air binding of RHR Pump B. Pump B was
immediately secured. The redundant RHR pump was started, and it too became
air bound and was secured. Because of anomalous performance of the
refueling, water level instrumentation, an operator entered containment to
read the tygon standpipe that had been installed temporarily to monitor
water level in the reactor vessel. The licensee concluded that suction to
the RHR pumps had been lost and started to raise the water level in the
reactor vessel. After level had increased 10 inches, an RHR pump was
restarted, but had to be secured because it still had inadequate suction
pressure. To provide pressure quickly and to increase level further, RHR
suction was transferred from the RCS to the refueling water storage tank.
The water level in the reactor vessel was raised an additional 2-1/2 feet.
Approximately 75 minutes after loss of decay heat removal, RHR Pump B was
vented and successfully returned to service. RHR Pump A was vented,
demonstrated to be operable, and deenergized. The reactor had been shut down
for approximately 100 days, and the increase in RCS temperature was
15F.
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Attachment 1
IN 86-101
December 12, 1986
Page 3 of 4
For Zion 2, the suction lines to the RHR pumps connect to a horizontal run
of RCS hot leg piping. The nozzles for the suction lines are located on the
underside of the hot leg piping and at a 45 angle from the bottom of
the line. The internal diameter of the hot legs is 29 inches, and the
internal diameter of the suction lines is 11 inches. When reactor water
level falls approximately 5.5 inches below the centerline of the hot leg,
uncovering of the RHR nozzle commences, and when the water level falls below
approximately 13.5 inches below the centerline, the RHR nozzle is completely
uncovered. During the event of December 14, 1985, vortexing started with
water level at 6 inches above the centerline with RHR flow at 3000 gpm.
A 10-inch line returns water from the RHR system to the RCS and is connected
to the top of one of the RCS cold legs. The water level sensing line for the
refueling water level transmitter is connected to a 4 inch line which is
connected to the same cold leg. Both nozzles are in the same vertical plane.
The 4-inch nozzle is located at 90 with respect to the 10-inch nozzle.
When the cold leg is partially filled as it was during this event, water
from the RHR return line impinges with appreciable force on the water
surface close to the nozzle for the 4-inch line. Because of possible dynamic
effects of this impingement, the operators believe that water level readings
from the refueling water level transmitter are inaccurate and erratic when
the water level in the reactor vessel is low. Furthermore, when the water
level in the reactor vessel is anywhere below the nominal midpoint of the
cold leg, the refueling water level instrument will indicate erroneously
that the water level is at the midpoint.
Notwithstanding these problems with the refueling water level
instrumentation, the tygon standpipe was not being continuously monitored
while the water level was low. Further, the operator did not know the
correlation of RHR flow rate and the water level for the onset of vortexing
at the suction of the RHR pumps.
Sequoyah 1
On October 9, 1985, Sequoyah 1, a Westinghouse designed reactor, was in cold
shutdown with the water level in the reactor vessel 4 inches below the
centers of the hot leg nozzles, RHR Train B in service for removal of decay
heat, and normal letdown and makeup out of service. The water level in the
reactor vessel had been lowered to facilitate plugging and eddy current
testing of tubes in a steam generator. During an evolution to put Train A in
service, RHR Pump A was started and then Pump B was secured. Running both
pumps simultaneously with low reactor vessel water level caused initiation
of vortexing and air binding in Pump A. The pump was secured immediately,
Pump B was restarted, and it operated normally. The alignment of Train A was
verified and the pump was vented . Pump B was secured, and Pump A was
restarted, became air bound, and was again secured. Pump B was restarted,
but this time it became air bound and was secured immediately. After
verifying that personnel were out of the steam generator, the water level in
the vessel was raised to the centerline of the hot legs by adding water to
the RCS from the RWST. Approximately 43 minutes after loss of decay heat
removal, Pump A was vented and returned to service. Pump B was vented,
demonstrated to be operable, and deenergized.
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Attachment 1
IN 86-101
December 12, 1986
Page 4 of 4
At Sequoyah 1, both RHR pumps take suction from the same hot leg (as they do
at San Onofre 2, Zion 2, and Catawba 1). The water level in the hot leg was
such that initially it would support operation of one RHR pump, but not both
pumps. Starting the second pump without first securing the operating pump
caused vortexing, air entrainment, and air binding of Pump A, which is
apparently more sensitive to this problem than Pump B. The procedure for
operating the RHR system with low water level in the reactor vessel did not
adequately reflect the relationship between RHR flow rate and water level
for the onset of vortexing in the suction line for the RHR pumps.
Catawba 1
On April 22, 1985, Catawba 1, a Westinghouse designed reactor, was in cold
shutdown with RHR Train A inoperable because of maintenance, and RHR Train B
in service to remove decay heat. Although one RHR train was inoperable, the
licensee started to lower the water level in the reactor vessel to
facilitate RCS pump seal maintenance. While draining was in progress,
erratic performance of RHR Pump B indicated that vortexing, air entrainment,
and air binding were occurring. The pump was secured; a charging pump was
aligned to take suction from the RWST; and the water level in the reactor
vessel was raised. Approximately 81 minutes after the first indication of
vortexing, RHR Pump B was vented and returned to service. Temperature of the
RCS peaked at 177 F.
For Catawba 1, the operating procedure for lowering water level in the
reactor vessel does limit RHR flow as a function of level, apparently to
preclude the onset of vortexing. However, the licensee believes that water
level information obtained from inaccurate instrumentation contributed to
complete loss of RHR flow. Further, the licensee incurred an increased risk
of loss of RHR flow by lowering water level with one train of RHR cooling
out of service. With the reactor in cold shutdown and the vessel partially
drained, a limiting condition for operation in the Technical Specifications,
for Catawba 1 requires that one RHR train be operating and that the other be
operable. Nevertheless, the operators concluded incorrectly that water level
could be lowered if corrective action had been initiated to comply with the
action statement for that limiting condition for operation.
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