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

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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 WARHR 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 for nuclear power reactors

Circuit Breaker Maintenance

Programs

99-12 Year 2000 Computer Systems 4/28/99 All holders of operating licenses

Readiness Audits or construction permits for nuclear

power plants

99-11 Incidents Involving the Use of 4/23/99 All medical use licensees

Radioactive Iodine-131

97-15, Sup 1 Reporting of Errors and 4/16/99 All holders of operating licenses

Changes in Large-Break/Small- for nuclear power reactors, except

Break Loss-of-Coolant Evaluation those who have permanently

Models of Fuel Vendors and cease operations and have

Compliance with 10 CFR 50.46(a)(3) certified that fuel has been

permanently removed from the

reactor

99-10 Degradation of Prestressing 4/13/99 All holders of operating licenses

Tendon Systems in Prestressed for nuclear power reactors

Concrete Containments

99-09 Problems Encountered When 3/24/99 All medical licensees authorized

Manually Editing Treatment Data to conduct high-dose-rate (HDR)

on The Nucletron Microselectron-HDR remote after loading

(New) Model 105.999 brachytherapy treatments

99-08 Urine Specimen Adulteration 4/1/99 All holders of operating licensees

for nuclear power reactors and

licensees authorized to possess

or use formula quantities of

strategic special nuclear material

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 F .

[3 /1/99 V. .

4/4I9

. .

1' /0g99 OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP I

NAME 2Jiiam 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 I Tech Editor l DRCH l-ii PDIV-1 lI

NAME CPetrone* 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