Information Notice 1990-05, Inter-System Discharge of Reactor Coolant
UK
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
January 29, 1990
NRC INFORMATION NOTICE NO. 90-05:
INTER-SYSTEM DISCHARGE OF REACTOR COOLANT
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose
This information notice is intended to. alert addressees to a potentially
significant problem in identifying and terminating reactor coolant system
leakage in operating modes 4 and 5. It is expected that licensees will
review the information for applicability to their facilities and consider
actions, as appropriate, to avoid similar problems. However, suggestions
contained in this information notice do not constitute NRC requirements;
therefore, no specific action or written response is required.
Description of Circumstances
On December 1, 1989, Braidwood Unit 1 experienced the unplanned inter-system
discharge of approximately 68,000 gallons of water. The discharge was caused
by the inadvertent opening of a residual heat removal (RHR) system suction
relief valve. The valve failed to reclose, allowing an open flow path from
the reactor vessel, through the RHR system, into the unit's two recycle hold-up
tanks (HUTs).
The unit, which had been in a refueling outage since September 2, 1989, was
heating up in operational mode 5, preparing to enter operational mode 4. The
plant was solid and in the process of drawing a bubble in the pressurizer. The
RHR train "A" pump was in operation and, although the "BO pump was not running, the "B" train was unisolated and available. The reactor coolant system (RCS)
was at a pressure of 350 psig .and a temperature of 1750F. Charging flow to the
vessel was being provided by the "A" charging pump. Pressurizer heaters were
on. The "B" charging pump was Isolated and tagged out of service.
(Technical
Specifications governing cold overpressure protection require that only one
charging pump be available. The other charging pump and the safety injection
pumps are required to be tagged out of service, with power supplies removed).
To protect against a pressure switch failure and the subsequent automatic
isolation of the RHR system, the train "A" RHR suction isolation valve was
open and tagged out of service.
90130126 Z #
IN 90-05 January 29, 1990 At 1:42 a.m., operators throttled the charging flow and maximized the letdown
flow in preparation for drawing a bubble in the pressurizer. The RCS pressure
was 404 psig and the pressurizer level was off scale, high. At 1:44 a.m., a
rapid reduction in the pressurizer level occurred, with the pressurizer level
off scale, low, at 1:52 a.m. Approximately 14,000 gallons of water drained
from the pressurizer and the pressurizer surge line; however, the reactor vessel
level instrumentation system indicated that the vessel level remained at 100
percent.
At 1:49 a.m., the charging flow was increased and the charging pump
suction was switched from the volume control tank to the refueling water storage
tank (RWST).
About 30 to 50 gallons of water were observed on the floor of the auxiliary
building in proximity to the RHR train "AN suction relief valve, leading plant
personnel to believe that this valve had lifted.
At 1:53 a.m., the letdown
flow was reduced to minimum and charging was maximized.
The RHR trains were
switched from "A" to EB", the "A" pump was stopped, and the isolation of the
"A" train was initiated. At 1:59 a.m., one of the two running reactor coolant
pumps (RCPs) was stopped because of low RCS pressure.
A second charging pump, NBN, was started following completion of the formal pro- cedure for tagout removal.
At 2:35 a.m., the "A RHR suction isolation valve
was returned to service and closed, completing the isolation of the "A" train
of the RHR system. The pressurizer level began to recover and the RCS pressure
increased slightly, giving operators the impression that the discharge had been
isolated.
The *B" charging pump was therefore secured at 2:45 a.m.
The pres- surizer level, however, did not recover. At 2:54 a.m., the ABN charging pump
was restarted. At 3:49 a.m., the inter-system discharge was terminated when
the RHR train WA" pump was started, the "B pump shut down, and the "8' train
was isolated. The level indication for the HUTs stabilized and the pressurizer
level began to recover at 3:52 a.m.
By 5:06 a.m., the pressurizer level had fully recovered and the unit was sta- bilized at 360 psi and 1750F. Approximately 68,000 gallons of water had been
discharged from the reactor vessel to the HUTs.
(The total amount of water
was composed of 14,000 gallons of initial pressurizer inventory and 54,000
gallons of makeup water).
Following the event, it was determined that the RHR MB" train suction relief
valve had lifted at 411 psi.
The lift setpoint for the valve should have been
450 psi. The valve should have reclosed on reducing pressure but failed to do
so.
The premature opening of the valve was attributed to the presence of foreign
material lodged between the valve spindle and the spindle guide. This foreign
material either prohibited the correct adjustment of the valve or affected the
valve's lift setpoint. The valve's failure to reclose was attributed to im- proper nozzle ring adjustment.
The reset pressure is strongly influenced by
the dynamic forces created by the nozzle ring.
If the ring is located too high
on the nozzle, it may result in an inadequate ventilation area just above the
nozzle.
Undesirable forces will develop which may cause a much lower reseat
pressure.
The water found near the RHR train "A" suction relief valve had leaked from
a weep hole on a relief valve in a radwaste evaporator line connected to the
IN 90-05 January 29, 1990 common discharge header of the train "A" and "B" suction relief valves.
Con- trary to original assumptions, there was no evidence that the OA" train suction
relief valve had lifted.
The root cause of the problem with the relief valve
on the evaporation line is under investigation but is thought to be unrelated
to the failure of the 'BM suction relief valve.
Hampering operators' efforts throughout this event was the lack of an appro- priate emergency operating procedure (EOP) to detect coolant leaks while in
operating modes 4 and 5. However, the operators were able to combine two
related abnormal operating procedures for guidance during this event. One
of the procedures is designed to locate system leaks while in modes 3 and 4.
The other provides guidance for the restoration of the RHR system following
its loss during conditions in which the reactor vessel inventory is at a
reduced level.
Discussion:
The event at Braidwood 1 is significant because it underscores the need to
have EOPs available for use in other than 'at power" operating modes. The
fact that over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> were required to locate the stuck-open valve, to
terminate the discharge, and to begin refilling the pressurizer highlights
the need to provide personnel with adequate tools to perform their tasks.
Relying on ad hoc procedures during significant events places an unnecessary
burden on operating personnel.
The lack of adequate EOPs could handicap the
most competent operators in their efforts to address significant operational
problems.
Also illustrated by this event Is the need for procedures to assure that
adequate RCS makeup capability and cooling options are available in a timely
fashion during shutdown.
The discharge through the stuck-open relief valve
exceeded the capability of a single charging pump.
Starting a second charging
pump required that formal procedures for tag removal be conducted.
This effort
necessitated a considerable amount of time, which may not be available should a
similar event occur while the RCS is at a higher temperature.
The severity of this event could have been increased if greater decay heat were
present in the reactor vessel or if a gross failure of the relief valve discharge
header had occurred. Greater decay heat would have increased the potential for
voiding in the core. Also, because the header discharges to the HUTs which are
located outside containment, a piping failure could have resulted in all or a
portion of the RCS water being discharged to the building floor. This event
would have necessitated a major cleanup effort and increased the potential for
personnel contamination.
If this event had occurred at one of the nuclear plants that has a single
suction line from the RCS to the RHR system, all shutdown cooling would
have been lost as a result of isolating the failed suction relief valve.
An alternate heat sink would likely have been required; however, in mode 5, an alternate heat sink may not be readily available.
IN 90-05 January 29, 1990 This information notice requires no specific action or written response.
If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate NRR project manager.
arl
E. ss, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts: Nick Fields, NRR
(301) 492-1173
Julian Hinds, RIII
(315) 388-5575 Attachment: List of Recently Issued NRC Information Notices
Attachment
January 29, 1990 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice-No..
Subject
-
Issuance
Issued to- el
90-04 Cracking of the Upper Shell- to-Transition Cone Girth
Welds in Steam Generators
1/26/90
All holders of OLs
or CPs for Westinghouse- designed and Combustion
Engineering-designed
nuclear power reactors.
90-03
90-02
90-01
89-90
89-89
89-88
89-87
89-45, Supp. 2
89-86 Malfunction of Borg-Warner
Bolted Bonnet Check Valves
Caused by Failure of the
Swing Arm
Potential Degradation of
Importance of Proper
Response to Self-Identified
Violations by Licensees
Pressurizer Safety Valve
Lift Setpoint Shift
Event Notification
Worksheets
Recent NRC-Sponsored
Testing of Motor-Operated
Valves
Disabling of Emergency
Diesel Generators. by
Their Neutral Ground-Fault
Protection Circuitry
Metalclad, Low-Voltage
Power Circuit Breakers
Refurbished with
Substandard Parts
Type HK Circuit Breakers
Missing Close Latch Anti- Shock Springs.
1/23/90
1/22/90
1/12/90
12/28/89
12/26/89
12/26/89
12/19/89
12/15/89
12/15/89
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
All holders of NRC
materials licenses.
All holders of OLs
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors.
OL = Operating License
CP = Construction Permit
IN 90-05 January 29, 1990 This information notice requires no specific action or written response.
If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate NRR project manager.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts:
Nick Fields, NRR
(301) 492-1173
Julian Hinds, RIII
(315) 388-5575 Attachment:
List of Recently Issued NRC Information Notices
- SEE PREVIOUS PAGE FOR CONCURRENCE
- EAB:NRR
NFields:db
1/12/90
- TECH:EDITOR *EAB:NRR
DCFischer
1/14/90
1/16/90
- C:EAB:NRR
CJHaughney
1/18/90
- C:OGCB:NRR
CHBerlinger
1/22 /90
Ross
11/.AY9O
.
-
IN 90-
January , 1990 No specific action or written response is required by this information
notice. If you have any questions about this matter, please contact one of
the technical contacts listed below or the Regional Administrator of the
appropriate regional office.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts:
Nick Fields, NRR
(301) 492-1173
Julian Hinds,RIII
(315) 388-5575 Attachment:
List of Recently Issued Information Notices
JJV
I'Ins
m
ofi w
EAB:NRR
TECH:EDITOR EAB:NRR
NFields:db
DCFischer
/ /,1-90
1 /*t/90
1/ i190
C: EB:NRR
CJHaughney
I As/90
coY
C:OGCB:NRR
CHBerlinger
I/.090
D:DOEA:NRR
CERossi
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/90