Information Notice 1999-14, Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrick

Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrick

ML031040444
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, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	05/05/1999
From:	Marsh L Division of Regulatory Improvement Programs
To:
References
IN-99-014, NUDOCS 9905070080
Download: ML031040444 (8)

UNITED STATES NUCLEAR REGULATORY

COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 May 5, 1999 NRC INFORMATION

NOTICE 99-14: UNANTICIPATED

REACTOR WATER DRAINDOWN AT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONE UNIT 2, AND FITZPATRICK

Addressees

All holders of licenses for nuclear power, test, and research reactors.

Purpose

The U.S. Nuclear Regulatory

Commission (NRC) is issuing this information

notice to alert addressees

to the potential

for personnel

errors during infrequently

performed

evolutions

that result in, or contribute

to, events such as the inadvertent

draining of water from the reactor vessel during shutdown operations.

It is expected that recipients

will review the information

for applicability

to their facilities

and consider actions, as appropriate, to prevent a similar occurrence.

However, suggestions

contained

in this information

notice are not NRC requirements;

therefore, no specific action or written response to this notice is required.DescriDtion

of Circumstances

Quad Cities Unit 2 On February 24, 1999, Quad Cities Unit 2 was in cold shutdown with reactor water temperature

at 131 'F and reactor water level at 80 inches indicated

level (normal level during operations

is 30 inches indicated

or 173 inches above the top of active fuel [TAF]). Core cooling was being maintained

in a band of 120 'F to 170 OF by the OA" loop of the shutdown cooling mode of the residual heat removal (RHR) system after being switched from the nB" loop at 12:32 a.m.During the switch over the licensee inadvertently

failed to close the OA RHR minimum flow valve as required by the procedure.

Sometime later operators

noted a decreasing

reactor water level and at about 1:02 a.m. secured the *2A RHR pump and isolated shutdown cooling. At 1:55 a.m. operators

restored the *2A' loop of shutdown cooling to the proper lineup and started the *2A RHR pump. Water level had decreased

to a minimum of about 45 inches indicated, and reactor water temperature

had risen to a maximum of about 163 OF. Forced circulation

of reactor vessel water using a reactor recirculation

pump remained in effect throughout

the event.On the basis of post event reviews, It appears that the minimum flow valve in the OA loop was left open because the nuclear station operator failed to ensure that the tasks were performed

in the sequence specified

in the operating

procedures.

The nuclear station operator who was (7008 PD(L H ort<<4qj-Oiif

qqos(J5 C7Ffcj ANW\\b

IN 99-14 May 5, 1999 directing

the evolution

from the control room gave the non-licensed

operator permission

to de-energize the breaker for the WA RHR minimum flow valve operator before the valve was taken to the required closed position.

De-energizing

the breaker also removed power to the valve position indicator

lights in the control room. Thus, when the nuclear station operator tried to verify that the valve was closed, there was no position indication

in the control room to make that verification.

The nuclear station operator made the incorrect

assumption

that the valve was already closed and moved to the next step in the procedure.

This failure to close the WAX RHR minimum flow valve opened a drain path from the reactor to the suppression

pool. To further complicate

the event, the operating

crew did not recognize

that there was any problem until approximately

10 minutes had passed and the water level had decreased

about 13 inches because of a misinterpretation

of causes of the level decrease.

After detecting

the decrease, the operating

crew was slow to react, which allowed the level to decrease another 20 inches before the operators

isolated shutdown cooling which terminated

the draindown.

The licensee estimated

that a total of 6000 to 7000 gallons was drained from the reactor to the suppression

pool.Operations

staff practices

including

poor communications, poor activity briefings

for high-risk activities, lack of effective

pre-shift

briefings, inadequate

supervision

of important

control room activities, inadequate

monitoring

of control room panels, and slow event response may have contributed

to the event. Although the unintended

loss of inventory

to the suppression

pool highlighted

significant

weaknesses

in plant operations, the safety significance

was minimized

by two features.

First, a reactor recirculation

pump remained in service throughout

the event which served to distribute

decay heat. Second, an automatic

isolation

of shutdown cooling would have occurred at 8 inches indicated

level which would have stopped the draining event.An indicated

water level of 8 inches corresponds

to approximately

151 inches of water level above the TAF in the reactor core.Arkansas Nuclear One Unit 2 On February 2, 1999, at Arkansas Nuclear One Unit 2, the operators

were draining the refueling

canal in preparation

for installing

the reactor vessel head. Refueling

was complete and steam generator

nozzle dams were installed.

The operators

were using the two low pressure safety injection (LPSI) pumps to drain the canal to the refueling

water storage tank;one pump also served as the shutdown cooling pump. The rate of draindown

was approximately

3.3 Inches per minute. When the water level reached 105 inches, the reactor operator noted that level started to lower rapidly. Operators

stopped one of the LPSI pumps and instructed

a local operator to close the isolation

valve to the refueling

water tank. This manually operated valve required 55 turns of the handwheel

to fully close. Within approximately

1.5 minutes, the reactor vessel level had dropped below the 65 inch level (where reduced inventory

begins) and continued

down to 56 inches before the valve could be fully closed. (Reference

zero on these level instruments

is the bottom of the hot leg, with mid-loop being defined at approximately

24 inches.) The average rate of level decrease between 105 IN 99-14 May 5, 1999 inches and 56 inches was approximately

33 inches per minute. At its lowest level, 56 inches indicated, there were still 93 inches of water above the TAF. Using the high pressure safety injection (HPSI) pump the operators

brought the level back up to 90 inches. The plant was in reduced inventory

operations (below 65 inches) for approximately

7 minutes. During the event the level remained well above the point where LPSI pump cavitation

would be expected.

The licensee concluded

that the safety significance

of the event was minimal because multiple sources of makeup water were available, redundant

mitigation

equipment

was available, and the operators

were quick to recognize

and respond to the event.On the basis of post event reviews, it was determined

that the procedure

used for draining down the refueling

canal was inadequate

in that it incorrectly

stated that the draindown

should be secured at the 90-inch level. The procedure

should have directed that the rate of draining be secured at the 106-inch level so that appropriate

precautions

could be taken before resuming the draindown.

These precautions

should have Included reminders

to the operating crew that below the 106-inch level the level will drop much more quickly due to the transition

of pumping from a large volume in the refueling

canal to a small volume In the reactor vessel.Therefore, in order to maintain control of the water level, the draindown

rate should be decreased

and an operator should be stationed

to directly monitor the level.Additional

factors that contributed

to this event include: the operators

received little specific training on this evolution;

the crew was inexperienced

in performing

this task; the task should have been classified

as an infrequent

task requiring

a more thorough briefing;

and, operators failed to station an operator in a position where he could directly monitor the water level in the refueling

canal. Instead they monitored

it remotely using a video camera that did not provide a clear picture of the water level.FitzPatrick

On December 2, 1998, at the James A. FitzPatrick

Nuclear Power Plant, the operators

were in the process of reassembling

the reactor following

refueling.

Operators

were controlling

the reactor vessel water level at 357 inches above TAF by adjusting

the water discharge

rate to compensate

for the constant input from the control rod drive cooling water system. While in this condition, the licensees

risk analysis requires that reactor vessel water level be monitored

using two independent

level indicators.

To meet this requirement, the licensee designated

a wide range indicator

which provided Indication

up to the top of the reactor vessel and an RHR interlock

level indicator

which provided indication

in the range from -150 inches to +200 Inches as the instruments

to be used during this evaluation.

In order for the wide-range

level Indicator

to remain available

with the reactor head removed, a temporary

standpipe

and fill funnel were used to replace a portion of the reference

leg. At the time of the event, the licensee was in the process of removing this temporary

standpipe

and reinstalling

the original reference

leg components.

As the water drained from the standpipe, it caused the wide-range

level indicator

to erroneously

show an increasing

water level. For a period of approximately

one hour the operators

in the control room, unaware that the ongoing maintenance

would cause an error in the indicated

water level, compensated

for the apparent increasing

level by increasing

the discharge

rate. This action had the effect of reducing the

IN 99-14 May 5, 1999 actual water level from 357 inches to 255 inches. During the same time period, the operators were also in the process of filling and venting the reactor feedwater

piping, which could have affected the reactor water level. Once the normal reference

leg piping had been reinstalled

and the reference

leg began to refill, the indicated

level decreased

from 357 inches to the actual level of 255 inches. The second level instrument, which does not come on-scale until the level goes below 200 inches, remained off-scale

high.When operators

discovered

the level discrepancy, they used a temporary

pressure gauge connected

to the reactor vessel low-point

tap to confirm the actual water level. After confirming

the accuracy of the wide-range

indicator, they restored the reactor vessel water level to 357 inches. The 100-inch error represented

approximately

14,000 gallons of water. The licensee determined

that the safety significance

of this event was low since the reactor was in cold shutdown with low decay heat and the reactor water level remained well above the TAF. In addition, the drain-down

would have been limited by an automatic

Isolation

of the draindown path, which would have occurred prior to vessel level reaching 177 Inches above the TAF.The licensee's

post event review identified:

weaknesses

in the operator's

knowledge

of the reactor assembly process; lack of explicit detail in the reactor assembly procedure;

and, weaknesses

in the plant risk assessment

process. Contrary to the assumption

that two designated

reactor water level indicators

were available, only one indicator, the wide-range

instrument, was available

in the range above 200 inches. When the reference

leg on the wide-range instrument

was disassembled

and drained, the one usable indicator

was rendered unavailable.

The second instrument

was pegged off-scale

high and remained that way throughout

the event because the level never dropped below 200 inches. A post event review by the licensee indicated

that other reactor water level instruments, remained operable during the event but, apparently

the operators

did not rely on these other instruments

or notice the discrepancy

between them and the wide range Indicator.

Proposed corrective

actions included procedural

enhancements

to ensure that reactor level instrumentation

credited by the outage risk assessment

remains available

during reactor disassembly

and reassembly.

Discussion

Personnel

errors appear to have caused, or contributed

to, these three inadvertent

reactor vessel draindown

events. The likelihood

of personnel

errors is dependent

upon the operators knowledge

of the task gained through previous experience

and training.

It is also dependent upon the quality of the procedures

used to perform the task, the level of supervision, the adequacy of pre-job briefings, fatigue, and distractions

resulting

from multiple tasks. In each of the events, the plant staff made errors during a seldom-performed

evolution.

Because it was a seldom-performed

evolution, more training, better pre-job briefings, closer supervision, and procedures

that contain more details than those for frequently

performed

activities

might have prevented

these events.

IN 99-14 May 5, 1999 This information

notice requires no specific action or written response.

If you have any questions

about the information

in this notice, please contact the technical

contact listed below, the appropriate

regional office, or the appropriate

Office of Nuclear Reactor Regulation (NRR)project manager.Ledyard B. Marsh, Chief Events Assessment, Generic Communications

And Non-Power

Reactors Branch Division of Regulatory

Improvement

Programs Office of Nuclear Reactor Regulation

Technical

contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdDRenrc.aov

REFERENCES:

NRC Integrated

Inspection

Report No. 50-333/98-08, issued February 10, 1999 (Accession

No.9902170348)

for the James A. FitzPatrick

Nuclear Power Plant for the period November 22, 1998, through January 10, 1999.Attachment:

List of Recently Issued NRC Information

Notices

~~ Attachment

1 IN 99-14 May 5, 1999 Page 1 of I LIST OF RECENTLY ISSUED NRC INFORMATION

NOTICES Information

Date of Notice No. Subject Issuance Issued to 99-13 Insiahts from NRR Inspections

4129199 All holders of operatina

licenses of Low-and Medium-Voltage

Circuit Breaker Maintenance

Programs for nuclear power reactors 99-12 Year 2000 Computer Systems Readiness

Audits Incidents

Involving

the Use of Radioactive

Iodine-131

4/28/99 4/23/99 All holders of operating

licenses or construction

permits for nuclear power plants All medical use licensees 99-11 97-15, Sup 1 Reporting

of Errors and 4/16/99 Changes in Large-Break/Small- Break Loss-of-Coolant

Evaluation

Models of Fuel Vendors and Compliance

with 10 CFR 50.46(a)(3)

All holders of operating

licenses for nuclear power reactors, except those who have permanently

cease operations

and have certified

that fuel has been permanently

removed from the reactor 99-10 99-09 Degradation

of Prestressing

4/13/99 Tendon Systems in Prestressed

Concrete Containments

Problems Encountered

When 3/24/99 Manually Editing Treatment

Data on The Nucletron

Microselectron-HDR (New) Model 105.999 Urine Specimen Adulteration

4/1/99 All holders of operating

licenses for nuclear power reactors All medical licensees

authorized

to conduct high-dose-rate (HDR)remote after loading brachytherapy

treatments

All holders of operating

licensees for nuclear power reactors and licensees

authorized

to possess or use formula quantities

of strategic

special nuclear material 99-08 OL = Operating

License CP = Construction

Permit

IN 99-xx April xx, 1999 Page 5of 5 This information

notice requires no specific action or written response.

If you have any questions

about the information

in this notice, please contact the technical

contact listed below, the appropriate

regional office, or the appropriate

office of Nuclear Reactor Regulation (NRR)Project Manager.Ledyard B. Marsh, Chief Events Assessment, Generic Communications

And Non-Power

Reactors Branch Division of Regulatory

Improvement

Programs Office of Nuclear Reactor Regulation

Technical

contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdRDanrc.aov

REFERENCES:

NRC Integrated

Inspection

Report No. 50-333198-08, issued February 10, 1999 (Accession

No.9902170348)

for the James A. FitzPatrick

Nuclear Power Plant for the period November 22, 1998, through January 10, 1999.Attachments:

1. List of Recently Issued NMSS Information

Notices 2. List of Recently Issued NRC Information

Notices DOCUMENT NAME: G:ICDPDRAININ\DRAIN.0B.WPD

To receive a copy of this document, Indicate In the box C=Copy w/o attachmentlenclosure

E=Copy with attachment/enclosure

N = No copy OFFICE PECB:DRIP

I Tech Editor l DRCH I PDIV-1 I NAME CPetrone I_ RGallo 1 MNolangfarP.

DATE V /0199 [3 /1/99 4 /4I9 1' /0g99 F .V. ...OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP

I NAME 2 Jiiam RPulsjier

LMarsh DATE lf/499 I1'/t 99 I /99 OFFICIAL RECORD COPY

IN 99-14 May 5, 1999 This information

notice requires no specific action or written response.

If you have any questions

about the information

in this notice, please contact the technical

contact listed below, the appropriate

regional office, or the appropriate

Office of Nuclear Reactor Regulation (NRR)project manager.[arig sjid by]Ledyard B. Marsh, Chief Events Assessment, Generic Communications

And Non-Power

Reactors Branch Division of Regulatory

Improvement

Programs Office of Nuclear Reactor Regulation

Technical

contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdr)ODnrc.gov

REFERENCES:

NRC Integrated

Inspection

Report No. 50-333/98-08, issued February 10, 1999 (Accession

No.9902170348)

for the James A. FitzPatrick

Nuclear Power Plant for the period November 22, 1998, through January 10, 1999.Attachment:

List of Recently Issued NRC Information

Notices DOCUMENT NAME: S:XDRPMSEC\99-14.IN

• See previous concurrence

To receive a copy of this document.

indicate in the box C=CoDv w/o attachment/enclosure

E=CoDv with attachment/enclosure

N = No coov OFFICE PECB:DRlIP

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BCalure* RGallo* MNolan*DATE 04/27/99 .3/15/99 _________04128199

= 04/27/99 1 ...OFFICE PDI-1 I PD111-2 C:PECB:DJRIP

I NAME JWilliams*

RPulsifer'

I-Marsh _ _ __ _DATE 04/27/9 .04/27/99 k,-u99 OFFICIAL RECORD COPY