Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
June 18, 1996
NRC INFORMATION NOTICE 96-37: INACCURATE REACTOR WATER LEVEL INDICATION
AND INADVERTENT DRAINDOWN DURING SHUTDOWN
Addressees
All pressurized water reactor facilities holding an operating license or a
construction permit.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to potential operational errors that may result in
the inadvertent loss of reactor coolant system (RCS) inventory during
refueling operations. It is expected that recipients will review this
information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems. However, suggestions contained in
this information notice are not NRC requirements; therefore, no specific
action or written response is required.
Description of Circumstances
On September 13, 1995, Surry Unit 1 was in a shutdown condition, cooled down,
and depressurized, preparatory to a refueling outage. A pressurizer power-
operated relief valve (PORV) and its block valve had been opened, connecting
the pressurizer to the pressurizer relief tank (PRT) which was pressurized
with nitrogen to 11 psig. The reactor head vent was open to the top of the
reactor water level indicating standpipe through the vent connection to the
PORV relief line (see Figure 1). The reactor coolant loop stop valves had
been closed, isolating the steam generators and reactor coolant pumps from
the reactor vessel and pressurizer. The water level in the reactor and
pressurizer had been lowered to slightly below the level of the reactor
vessel flange (elevation 18 feet) so that the vessel head studs could be
de-tensioned. The pressurizer was empty, and the reactor coolant piping was
full to part way up the surge line. The isolated portion of the "A" reactor
coolant loop was being drained.
To install the cavity seal ring so that the cavity could be flooded up to
permit lifting the reactor head, the reactor head vent was closed and the
ventline spoolpiece was disconnected. After the seal ring was in place, the
spoolpiece was re-connected but the reactor head vent valve was not reopened
(no step in the procedure being used called for opening the valve). This
resulted in loss of function of the only reactor water level indication
available while the reactor vessel head was still installed. [The reactor
vessel level indication system (RVLIS) had been used during the earlier
draindown process, but had been disconnected, as usual, in anticipation of the
removal of the reactor vessel heads.]
9606120154 . IN 96-37
June 18, 1996
Page 2 of 3
The PRT nitrogen pressure was gradually being reduced. An operator saw the
standpipe indicated level increase as the gas bubble trapped in the reactor
vessel head expanded, forcing water out of the reactor and up the surge line
and standpipe. Unaware of the closed head vent and believing the standpipe
level indication, the operator increased letdown from the reactor coolant
system cold leg piping to maintain the indicated 18-foot level. As the
increased letdown raised the water level in the volume control tank (VCT), the
automatic VCT level control bypassed some of the letdown from the VCT to a
holdup tank. This effectively reduced the inventory in the reactor coolant
system by approximately 4500 gallons over a period of approximately 5 hours.
When the PRT pressure was subsequently reduced to atmospheric pressure and the
reactor vessel head studs were de-tensioned, the vessel internals spring
loading lifted the head enough to relieve the reactor internal pressure, which
caused a sudden drop in indicated level in the standpipe. The indicated level
dropped from elevation 18 feet to elevation 13.3 feet. The operator
immediately took action to restore the level to elevation 18 feet.
The capability to maintain decay heat removal was not reduced during this
event because adequate suction head remained on the residual heat removal
(RHR) pumps; the 13.3-foot water level is 1.5 feet above mid-loop and more
than 6 feet above the core. There was never an indication of reduction in RHR
flow during the event. If the gas bubble had continued to expand, the reactor
vessel water level would not have dropped lower than just below the top of the
reactor coolant system hot leg pipe; at that point the gas bubble would vent
through the hot leg pipe into the pressurizer and into the PRT, equalizing
pressure between the PRT and vessel atmosphere. The standpipe indicated level
would then have been correct.
Discussion
Factors contributing to this event involved apparent weaknesses in operational
procedures:
1. The procedure for reconnecting the standpipe did not require re-opening
of the vessel head vent valve.
2. Operability of the water level indication system was not determined
after it had been disabled when the cavity seal ring was to be
installed.
3. The relationship between the reactor vessel head vent and the capability
of the standpipe to indicate water level accurately was apparently not
clear to the control room operators.
4. With the loop stop valves closed, there was no apparent reason why
standpipe level should rise. This relationship apparently was not clear
to the operators.
. IN 96-37
June 18, 1996
Page 3 of 3
5. Although tank levels were noted three times a day, inventory balances
were not determined from the data.
Surry and a few other pressurized water reactor (PWR) plants have reactor
coolant loop stop valves. Most PWRs do not. Should a similar event occur at
a plant without these valves, the course of the event would not be
significantly different from that which occurred at Surry. The water in the
steam generator tubes would drain through the reactor coolant piping into the
reactor vessel as the gas pressure in the PRT was reduced, extending the time
required for reactor vessel water level to reach the top of the inside of
the hot leg pipe. An event similar to this Surry event occurred at the
Sequoyah Nuclear Plant, which does not have loop stop valves. The Sequoyah
event was the subject of NRC Information Notice 94-36, "Undetected
Accumulation of Gas in Reactor Coolant System," issued May 24, 1994.
Further information concerning the Surry event may be found in NRC Inspection
Report 50-280/95-20 and 50-281/95-20.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
any of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
signed by
Brian K. Grimes, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Muhammad M. Razzaque, NRR
(301) 415-2882
Internet:[email protected]
Morris Branch, RII
(804) 357-2101
Internet:[email protected]
Robert A. Benedict, NRR
(301) 415-1157
Internet:[email protected]
Attachments:
1. Figure 1. Reactor Vessel Head Vent and Standpipe Diagram
2. List of Recently Issued NRC Information Notices
Note: Attachments not included in this version
