Information Notice 1990-05, Inter-System Discharge of Reactor Coolant: Difference between revisions

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{{#Wiki_filter:UK UNITED STATES NUCLEAR REGULATORY
{{#Wiki_filter:UK


COMMISSION
UNITED STATES


===OFFICE OF NUCLEAR REACTOR REGULATION===
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555 January 29, 1990 NRC INFORMATION


NOTICE NO. 90-05: INTER-SYSTEM
OFFICE OF NUCLEAR REACTOR REGULATION


DISCHARGE
WASHINGTON, D.C. 20555 January 29, 1990
 
NRC INFORMATION NOTICE NO. 90-05:    INTER-SYSTEM DISCHARGE OF REACTOR COOLANT
OF REACTOR COOLANT


==Addressees==
==Addressees==
:
:
All holders of operating
All holders of operating licenses or construction permits for nuclear power
 
licenses or construction


permits for nuclear power reactors.
reactors.


==Purpose==
==Purpose==
: This information
:
This information notice is intended to. alert addressees to a potentially


notice is intended to. alert addressees
system


to a potentially
significant problem in identifying and terminating reactor coolant will


significant
leakage in operating modes 4 and 5. It is expected that licenseesconsider


problem in identifying
review the information for applicability to their facilities and


and terminating
actions, as appropriate, to avoid similar problems. However, suggestions


reactor coolant system leakage in operating
contained in this information notice do not constitute NRC requirements;
therefore, no specific action or written response is required.


modes 4 and 5. It is expected that licensees
==Description of Circumstances==
:
On December 1, 1989, Braidwood Unit 1 experienced the unplanned inter-system


will review the information
discharge of approximately 68,000 gallons of water. The discharge was caused


for applicability
by the inadvertent opening of a residual heat removal (RHR) system suction


to their facilities
relief valve. The valve failed to reclose, allowing an open flow path from


and consider actions, as appropriate, to avoid similar problems.
the reactor vessel, through the RHR system, into the unit's two recycle hold-up


===However, suggestions===
tanks (HUTs).
contained


in this information
The unit, which had been in a refueling outage since September 2, 1989,4.wasThe


notice do not constitute
heating up in operational mode 5, preparing to enter operational mode


NRC requirements;
plant was solid and in the process of drawing a bubble in the pressurizer. The
therefore, no specific action or written response is required.Description


of Circumstances:
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)
On December 1, 1989, Braidwood
                                                        0
was at a pressure of 350 psig .and a temperature of 175 F. Charging flow to the


Unit 1 experienced
vessel was being provided by the "A" charging pump. Pressurizer heaters were


the unplanned
on. The "B" charging pump was Isolated and tagged out of service. (Technical


inter-system
Specifications governing cold overpressure protection require that only one


discharge
injection


of approximately
charging pump be available. The other charging pump and the safety removed).


68,000 gallons of water. The discharge
pumps are required to be tagged out  of service, with power supplies


was caused by the inadvertent
To protect against a pressure switch failure and the subsequent automatic


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
isolation of the RHR system, the train "A" RHR suction isolation valve was


outage since September
open and tagged out of service.


2, 1989, was heating up in operational
90130126 Z#


mode 5, preparing
IN 90-05 January 29, 1990 At 1:42 a.m., operators throttled the charging flow and maximized


to enter operational
flow in preparation for drawing a bubble in the pressurizer. The the letdown


mode 4. The plant was solid and in the process of drawing a bubble in the pressurizer.
was 404 psig and the pressurizer level was off scale, high. At        RCS pressure


The RHR train "A" pump was in operation
1:44 a.m., a


and, although the "BO pump was not running, the "B" train was unisolated
rapid reduction in the pressurizer level occurred, with the pressurizer


and available.
off scale, low, at 1:52 a.m. Approximately 14,000 gallons of water            level


The reactor coolant system (RCS)was at a pressure of 350 psig .and a temperature
from the pressurizer and the pressurizer surge line; however, the      drained


of 175 0 F. Charging flow to the vessel was being provided by the "A" charging pump. Pressurizer
level instrumentation system indicated that the vessel level remainedreactor vessel


heaters were on. The "B" charging pump was Isolated and tagged out of service. (Technical
percent. At 1:49 a.m., the charging flow was increased and the            at 100
  suction was switched from the volume control tank to the refueling charging    pump


Specifications
tank (RWST).                                                          water    storage


governing
About 30 to 50 gallons of water were observed on the floor of the


cold overpressure
building in proximity to the RHR train "AN suction relief valve, auxiliary


protection
personnel to believe that this valve had lifted. At 1:53 a.m., leading plant


require that only one charging pump be available.
flow was reduced to minimum and charging was maximized. The RHR the letdown


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
switched from "A" to EB", the "A" pump was stopped, and the isolationtrains were


automatic isolation
"A" train was initiated. At 1:59 a.m., one of the two running              of the


of the RHR system, the train "A" RHR suction isolation
pumps (RCPs) was stopped because of low RCS pressure.            reactor    coolant


valve was open and tagged out of service.90130126 Z #
A second charging pump, NBN, was started following completion of
IN 90-05 January 29, 1990 At 1:42 a.m., operators


throttled
the formal pro- cedure for tagout removal. At 2:35 a.m., the "A RHR suction isolation


the charging flow and maximized
was returned to service and closed, completing the isolation of              valve


the letdown flow in preparation
of the RHR system. The pressurizer level began to recover and the  the  "A"  train


for drawing a bubble in the pressurizer.
increased slightly, giving operators the impression that the discharge RCS  pressure


The RCS pressure was 404 psig and the pressurizer
isolated. The *B" charging pump was therefore secured at 2:45                had been


level was off scale, high. At 1:44 a.m., a rapid reduction
surizer level, however, did not recover. At 2:54 a.m., the ABN  a.m.    The  pres- charging


in the pressurizer
was restarted. At 3:49 a.m., the inter-system discharge was terminated pump


level occurred, with the pressurizer
the RHR train WA" pump was started, the "B pump shut down, and                when


level off scale, low, at 1:52 a.m. Approximately
the "8' train


14,000 gallons of water drained from the pressurizer
was isolated. The level indication for the HUTs stabilized and


and the pressurizer
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
level began to recover at 3:52 a.m.


and the pressurizer
By 5:06 a.m., the pressurizer level had fully recovered and the


level began to recover at 3:52 a.m.By 5:06 a.m., the pressurizer
unit was sta- bilized at 360 psi and 1750F. Approximately 68,000 gallons of water


level had fully recovered
discharged from the reactor vessel to the HUTs. (The total amount          had been


and the unit was sta-bilized at 360 psi and 1750F. Approximately
was composed of 14,000 gallons of initial pressurizer inventory        of  water


68,000 gallons of water had been discharged
gallons of makeup water).                                          and  54,000
Following the event, it was determined that the RHR MB" train


from the reactor vessel to the HUTs. (The total amount of water was composed of 14,000 gallons of initial pressurizer
suction


inventory
valve had lifted at 411 psi. The lift setpoint for the valve should relief


and 54,000 gallons of makeup water).Following
450 psi. The valve should have reclosed on reducing pressure but          have been


the event, it was determined
so. The premature opening of the valve was attributed to the          failed    to do


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
material lodged between the valve spindle and the spindle guide. presence    of  foreign


opening of the valve was attributed
material either prohibited the correct adjustment of the valve         This  foreign


to the presence of foreign material lodged between the valve spindle and the spindle guide. This foreign material either prohibited
or affected the


the correct adjustment
valve's lift setpoint. The valve's failure to reclose was attributed


of the valve or affected the valve's lift setpoint.
proper nozzle ring adjustment. The reset pressure is strongly                to im- influenced by


The valve's failure to reclose was attributed
the dynamic forces created by the nozzle ring. If the ring is


to im-proper nozzle ring adjustment.
located too high


The reset pressure is strongly influenced
on the nozzle, it may result in an inadequate ventilation area


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
nozzle. Undesirable forces will develop which may cause a much    just    above the


ventilation
pressure.                                                          lower    reseat


area just above the nozzle. Undesirable
The water found near the RHR train "A" suction relief valve had


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
leaked from


line connected
a weep hole on a relief valve in a radwaste evaporator line connected


to the
to the


IN 90-05 January 29, 1990 common discharge
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
 
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 hours 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
relief valve had lifted. The root cause of the problem with the relief valve


burden on operating
on the evaporation line is under investigation but is thought to be unrelated


personnel.
to the failure of the 'BM suction relief valve.


The lack of adequate EOPs could handicap the most competent
Hampering operators' efforts throughout this event was the lack of an appro- priate emergency operating procedure (EOP) to detect coolant leaks while in


operators
operating modes 4 and 5. However, the operators were able to combine two


in their efforts to address significant
related abnormal operating procedures for guidance during this event. One


operational
of the procedures is designed to locate system leaks while in modes 3 and 4.


problems.Also illustrated
The other provides guidance for the restoration of the RHR system following


by this event Is the need for procedures
its loss during conditions in which the reactor vessel inventory is at a


to assure that adequate RCS makeup capability
reduced level.


and cooling options are available
Discussion:
The event at Braidwood 1 is significant because it underscores the need to


in a timely fashion during shutdown.
have EOPs available for use in other than 'at power" operating modes. The


The discharge
fact that over 2 hours were required to locate the stuck-open valve, to


through the stuck-open
terminate the discharge, and to begin refilling the pressurizer highlights


relief valve exceeded the capability
the need to provide personnel with adequate tools to perform their tasks.


of a single charging pump. Starting a second charging pump required that formal procedures
Relying on ad hoc procedures during significant events places an unnecessary


for tag removal be conducted.
burden on operating personnel. The lack of adequate EOPs could handicap the


This effort necessitated
most competent operators in their efforts to address significant operational


a considerable
problems.


amount of time, which may not be available
Also illustrated by this event Is the need for procedures to assure that


should a similar event occur while the RCS is at a higher temperature.
adequate RCS makeup capability and cooling options are available in a timely


The severity of this event could have been increased
fashion during shutdown. The discharge through the stuck-open relief valve


if greater decay heat were present in the reactor vessel or if a gross failure of the relief valve discharge header had occurred.
exceeded the capability of a single charging pump. Starting a second charging


Greater decay heat would have increased
pump required that formal procedures for tag removal be conducted. This effort


the potential
necessitated a considerable amount of time, which may not be available should a


for voiding in the core. Also, because the header discharges
similar event occur while the RCS is at a higher temperature.


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
The severity of this event could have been increased if greater decay heat were


to the building floor. This event would have necessitated
present in the reactor vessel or if a gross failure of the relief valve discharge


a major cleanup effort and increased
header had occurred. Greater decay heat would have increased the potential for


the potential
voiding in the core. Also, because the header discharges to the HUTs which are


for personnel
located outside containment, a piping failure could have resulted in all or a


contamination.
portion of the RCS water being discharged to the building floor. This event


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
would have necessitated a major cleanup effort and increased the potential for


the failed suction relief valve.An alternate
personnel contamination.


heat sink would likely have been required;
If this event had  occurred at one of the nuclear plants that has a single
however, in mode 5, an alternate


heat sink may not be readily available.
suction line from  the RCS to the RHR system, all shutdown cooling would


IN 90-05 January 29, 1990 This information
have been lost as  a result of isolating the failed suction relief valve.


notice requires no specific action or written response.
An alternate heat  sink would likely have been required; however, in mode 5, an alternate heat  sink may not be readily available.


If you have any questions
IN 90-05 January 29, 1990 This information notice requires no specific action or written response. If


about the information
you have any questions about the information in this notice, please contact


in this notice, please contact one of the technical
one of the technical contacts listed below or the appropriate NRR project manager.


contacts listed below or the appropriate
arl  E. ss, Director


NRR project manager.arl E. ss, Director Division of Operational
Division of Operational Events Assessment


===Events Assessment===
Office of Nuclear Reactor Regulation
Office of Nuclear Reactor Regulation


Technical
Technical Contacts:  Nick Fields, NRR


Contacts:
(301) 492-1173 Julian Hinds, RIII
Nick Fields, NRR (301) 492-1173 Julian Hinds, RIII (315) 388-5575 Attachment:
List of Recently Issued NRC Information


Notices
(315) 388-5575 Attachment:  List of Recently Issued NRC Information Notices


Attachment
Attachment


IN 90-05 January 29, 1990 LIST OF RECENTLY ISSUED NRC INFORMATION
IN 90-05 January 29, 1990 LIST OF RECENTLY ISSUED


NOTICES Information
NRC INFORMATION NOTICES


Date of Notice-No..
Information                                    Date of


Subject -Issuance Issued to-el 90-04 Cracking of the Upper Shell-to-Transition
Notice-No..    Subject                   -   Issuance   Issued to- el 90-04           Cracking of the Upper Shell-   1/26/90    All holders of OLs


===Cone Girth Welds in Steam Generators===
to-Transition Cone Girth                 or CPs for Westinghouse- Welds in Steam Generators                designed and Combustion
1/26/90 All holders of OLs or CPs for Westinghouse- designed and Combustion


Engineering-designed
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
nuclear power reactors.


of Borg-Warner
90-03          Malfunction of Borg-Warner     1/23/90    All holders of OLs


Bolted Bonnet Check Valves Caused by Failure of the Swing Arm Potential
Bolted Bonnet Check Valves               or CPs for nuclear


Degradation
Caused by Failure of the                  power reactors.


of Secondary
Swing Arm


Containment
90-02          Potential Degradation of        1/22/90  All holders of OLs


Importance
Secondary Containment                      or CPs for BWRs.


of Proper Response to Self-Identified
90-01          Importance of Proper           1/12/90    All holders of NRC


Violations
Response to Self-Identified                materials licenses.


by Licensees Pressurizer
Violations by Licensees


===Safety Valve Lift Setpoint Shift Event Notification===
89-90          Pressurizer Safety Valve       12/28/89  All holders of OLs
Worksheets


Recent NRC-Sponsored
Lift Setpoint Shift                      or CPs for PWRs.


Testing of Motor-Operated
89-89          Event Notification            12/26/89  All holders of OLs


Valves Disabling
Worksheets                                or CPs for nuclear


of Emergency Diesel Generators.
power reactors.


by Their Neutral Ground-Fault
89-88          Recent NRC-Sponsored          12/26/89  All holders of OLs


Protection
Testing of Motor-Operated                  or CPs for nuclear


Circuitry Metalclad, Low-Voltage
Valves                                    power reactors.


Power Circuit Breakers Refurbished
89-87          Disabling of Emergency        12/19/89    All holders of OLs


with Substandard
Diesel Generators. by                      or CPs for nuclear


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 or CPs for BWRs.All holders of NRC materials
Their Neutral Ground-Fault                power reactors.


licenses.All holders of OLs or CPs for PWRs.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
Protection Circuitry


License CP = Construction
89-45,          Metalclad, Low-Voltage          12/15/89  All holders of OLs


Permit
Supp. 2        Power Circuit Breakers                    or CPs for nuclear


IN 90-05 January 29, 1990 This information
Refurbished with                          power reactors.


notice requires no specific action or written response.
Substandard Parts


If you have any questions
89-86          Type HK Circuit Breakers        12/15/89  All holders of OLs


about the information
Missing Close Latch Anti-                  or CPs for nuclear


in this notice, please contact one of the technical
Shock Springs.                            power reactors.


contacts listed below or the appropriate
OL = Operating License


NRR project manager.Charles E. Rossi, Director Division of Operational
CP = Construction Permit


===Events Assessment===
IN 90-05 January 29, 1990 This information notice requires no specific action or written response. If
Office of Nuclear Reactor Regulation


Technical
you have any questions about the information in this notice, please contact


Contacts: Nick Fields, NRR (301) 492-1173 Julian Hinds, RIII (315) 388-5575 Attachment:
one of the technical contacts listed below or the appropriate NRR project manager.
List of Recently Issued NRC Information


Notices*SEE PREVIOUS PAGE FOR CONCURRENCE
Charles E. Rossi, Director


*EAB:NRR NFields:db
Division of Operational Events Assessment


1/12/90*TECH:EDITOR
Office of Nuclear Reactor Regulation


*EAB:NRR DCFischer 1/14/90 1/16/90*C:EAB:NRR
Technical Contacts:   Nick Fields, NRR


CJHaughney
(301) 492-1173 Julian Hinds, RIII


1/18/90*C:OGCB:NRR
(315) 388-5575 Attachment: List of Recently Issued NRC Information Notices


CHBerlinger
*SEE PREVIOUS PAGE FOR CONCURRENCE


1/22 /90 Ross 11/.AY9O
*EAB:NRR    *TECH:EDITOR  *EAB:NRR    *C:EAB:NRR  *C:OGCB:NRR


.-;IN 90-January , 1990 No specific action or written response is required by this information
NFields:db                  DCFischer    CJHaughney  CHBerlinger      Ross


notice. If you have any questions
1/12/90      1/14/90      1/16/90      1/18/90      1/22 /90      11/.AY9O


about this matter, please contact one of the technical
. - ;                                                              IN 90-
                                                                  January , 1990 No specific action or written response is required by this information


contacts listed below or the Regional Administrator
notice. If you have any questions about this matter, please contact one of


of the appropriate
the technical contacts listed below or the Regional Administrator of the


regional office.Charles E. Rossi, Director Division of Operational
appropriate regional office.


===Events Assessment===
Charles E. Rossi, Director
Office of Nuclear Reactor Regulation


Technical
Division of Operational Events Assessment


Contacts: Nick Fields, NRR (301) 492-1173 Julian Hinds,RIII
Office of Nuclear Reactor Regulation


(315) 388-5575 Attachment:
Technical Contacts:       Nick Fields, NRR
List of Recently Issued Information


Notices JJV I'Ins m ofi w EAB:NRR TECH:EDITOR
(301) 492-1173 Julian Hinds,RIII


EAB:NRR NFields:db
(315) 388-5575 Attachment:     List of Recently Issued Information Notices


DCFischer/ /,1-90 1 /*t/90 1/ i190 C: EB:NRR CJHaughney
JJV        m


I As/90 coY C:OGCB:NRR
I'Ins  ofi w                              coY


CHBerlinger
EAB:NRR      TECH:EDITOR EAB:NRR        C: EB:NRR  C:OGCB:NRR      D:DOEA:NRR


I/.090 D:DOEA:NRR
NFields:db                  DCFischer  CJHaughney CHBerlinger    CERossi


CERossi/ /90}}
/ /,1-90        1 /*t/90    1/ i190      I As/90    I/.090          / /90}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Latest revision as of 04:02, 24 November 2019

Inter-System Discharge of Reactor Coolant
ML031130342
Person / Time
Site: Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date: 01/29/1990
From: Rossi C
Office of Nuclear Reactor Regulation
To:
References
IN-90-005, NUDOCS 9001230126
Download: ML031130342 (8)
v • d • e


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

system

significant problem in identifying and terminating reactor coolant will

leakage in operating modes 4 and 5. It is expected that licenseesconsider

review the information for applicability to their facilities and

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, 1989,4.wasThe

heating up in operational mode 5, preparing to enter operational mode

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)

0

was at a pressure of 350 psig .and a temperature of 175 F. 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

injection

charging pump be available. The other charging pump and the safety removed).

pumps are required to be tagged out of service, with power supplies

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

flow in preparation for drawing a bubble in the pressurizer. The the letdown

was 404 psig and the pressurizer level was off scale, high. At RCS pressure

1:44 a.m., a

rapid reduction in the pressurizer level occurred, with the pressurizer

off scale, low, at 1:52 a.m. Approximately 14,000 gallons of water level

from the pressurizer and the pressurizer surge line; however, the drained

level instrumentation system indicated that the vessel level remainedreactor vessel

percent. At 1:49 a.m., the charging flow was increased and the at 100

suction was switched from the volume control tank to the refueling charging pump

tank (RWST). water storage

About 30 to 50 gallons of water were observed on the floor of the

building in proximity to the RHR train "AN suction relief valve, auxiliary

personnel to believe that this valve had lifted. At 1:53 a.m., leading plant

flow was reduced to minimum and charging was maximized. The RHR the letdown

switched from "A" to EB", the "A" pump was stopped, and the isolationtrains were

"A" train was initiated. At 1:59 a.m., one of the two running of the

pumps (RCPs) was stopped because of low RCS pressure. reactor coolant

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

was returned to service and closed, completing the isolation of valve

of the RHR system. The pressurizer level began to recover and the the "A" train

increased slightly, giving operators the impression that the discharge RCS pressure

isolated. The *B" charging pump was therefore secured at 2:45 had been

surizer level, however, did not recover. At 2:54 a.m., the ABN a.m. The pres- charging

was restarted. At 3:49 a.m., the inter-system discharge was terminated pump

the RHR train WA" pump was started, the "B pump shut down, and when

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

discharged from the reactor vessel to the HUTs. (The total amount had been

was composed of 14,000 gallons of initial pressurizer inventory of water

gallons of makeup water). and 54,000

Following the event, it was determined that the RHR MB" train

suction

valve had lifted at 411 psi. The lift setpoint for the valve should relief

450 psi. The valve should have reclosed on reducing pressure but have been

so. The premature opening of the valve was attributed to the failed to do

material lodged between the valve spindle and the spindle guide. presence of foreign

material either prohibited the correct adjustment of the valve This foreign

or affected the

valve's lift setpoint. The valve's failure to reclose was attributed

proper nozzle ring adjustment. The reset pressure is strongly to im- 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

nozzle. Undesirable forces will develop which may cause a much just above the

pressure. lower reseat

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

IN 90-05 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- 1/26/90 All holders of OLs

to-Transition Cone Girth or CPs for Westinghouse- Welds in Steam Generators designed and Combustion

Engineering-designed

nuclear power reactors.

90-03 Malfunction of Borg-Warner 1/23/90 All holders of OLs

Bolted Bonnet Check Valves or CPs for nuclear

Caused by Failure of the power reactors.

Swing Arm

90-02 Potential Degradation of 1/22/90 All holders of OLs

Secondary Containment or CPs for BWRs.

90-01 Importance of Proper 1/12/90 All holders of NRC

Response to Self-Identified materials licenses.

Violations by Licensees

89-90 Pressurizer Safety Valve 12/28/89 All holders of OLs

Lift Setpoint Shift or CPs for PWRs.

89-89 Event Notification 12/26/89 All holders of OLs

Worksheets or CPs for nuclear

power reactors.

89-88 Recent NRC-Sponsored 12/26/89 All holders of OLs

Testing of Motor-Operated or CPs for nuclear

Valves power reactors.

89-87 Disabling of Emergency 12/19/89 All holders of OLs

Diesel Generators. by or CPs for nuclear

Their Neutral Ground-Fault power reactors.

Protection Circuitry

89-45, Metalclad, Low-Voltage 12/15/89 All holders of OLs

Supp. 2 Power Circuit Breakers or CPs for nuclear

Refurbished with power reactors.

Substandard Parts

89-86 Type HK Circuit Breakers 12/15/89 All holders of OLs

Missing Close Latch Anti- or CPs for nuclear

Shock Springs. 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 *TECH:EDITOR *EAB:NRR *C:EAB:NRR *C:OGCB:NRR

NFields:db DCFischer CJHaughney CHBerlinger Ross

1/12/90 1/14/90 1/16/90 1/18/90 1/22 /90 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 m

I'Ins ofi w coY

EAB:NRR TECH:EDITOR EAB:NRR C: EB:NRR C:OGCB:NRR D:DOEA:NRR

NFields:db DCFischer CJHaughney CHBerlinger CERossi

/ /,1-90 1 /*t/90 1/ i190 I As/90 I/.090 / /90


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