Information Notice 1990-05, Inter-System Discharge of Reactor Coolant

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 E Office of Nuclear Reactor Regulation
To:
References
IN-90-005, NUDOCS 9001230126
Download: ML031130342 (8)

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 175 0 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 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

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-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 Secondary

Containment

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 or CPs for BWRs.All holders of NRC materials

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

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/ /90