05000424/LER-2003-001, Debris in Containment Could Have Resulted in Safety System Loss of Function

Debris in Containment Could Have Resulted in Safety System Loss of Function

ML032580413
Person / Time
Site:	Vogtle
Issue date:	09/11/2003
From:	Gasser J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-03-1819 LER 03-001-00
Download: ML032580413 (5)

LER-2003-001, Debris in Containment Could Have Resulted in Safety System Loss of Function

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
4242003001R00 - NRC Website

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7r Jeffrey T Gasser Southern Nuclear Vice President Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.7721 Fax 205.992.0403 S

SOUTHERNN September 11, 2003 COMPANY Energy to Serve Your We rid Docket No.:

50-424 NL-03-1819 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant Licensee Event Report 1-2003-001 Debris in Containment Could Have Resulted in Safety System Loss of Function Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50.73, Southern Nuclear Operating Company hereby submits a Vogtle Electric Generating Plant Licensee Event Report (LER) for a condition that occurred on April 6, 2002.

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely, T. Gasser JTG/NJS/daj Enclosure: LER 1-2003-001 cc:

Southern Nuclear Operating Company Mr. J. D. Woodard, Executive Vice President Mr. W. F. Kitchens, General Manager - Plant Vogtle Mr. M. Sheibani, Engineering Supervisor - Plant Vogtle Document Services RTYPE: CVC7000 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. F. Rinaldi, NRR Project Manager - Vogtle Mr. J. Zeiler, Senior Resident Inspector - Vogtle

Abstract

On July 28, 2003, an evaluation was completed that determined a safety system functional failure could have occurred on April 6 & 7,2002, following the IRIO refueling outage. Specifically, it was determined that the containment building held debris in an amount adequate to block the containment sump screens, which could result in inadequate net positive suction head and possible failure of the residual heat removal (RHR) pumps to perform at design limiting conditions. This may have occurred following a design basis accident had these pumps been called on to operate and draw suction from the containment sumps. The condition existed for a period of 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> and 12 minutes while the unit was subcritical, until a containment cleanliness walkdown removed enough of the debris to reduce the cross-sectional area of the sump screen blockage to an acceptable level. The July 28, 2003, evaluation included expanded insulation destruction due to missing insulation jacketing which increased the insulation debris contributing to sump blockage.

The cause of this event was inadequate removal of debris from the Unit 1 containment building prior to Mode 4 (hot shutdown) heat-up. Debris was removed prior to the unit's return to power operations.U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE (3)

YER I SEQUJENTIA0L I REVISION3 I.

l YEAR l tEQY~s4NR71ALYlRNUMBER Vogtle Electric Generating Plant - Unit 1 05000-424 2003 -- 001

-- 00 2 OF 4 TEXT fif more space is required, use additional copies of NRC Form 366A) (17 A. REQUIREMENT FOR REPORT This event is reportable per 10 CFR 50.73 (aX2)(v)(B) because a condition existed that could have prevented the fulfillment of the safety function of a system needed to remove residual heat. It is also reportable per 10 CFR 50.73 (a)(2)(i)(B) because the unit operated in a condition prohibited by the Technical Specifications when a surveillance task was inadequately performed.

B. UNIT STATUS AT TIME OF EVENT At the time of this event on April 6, 2002, Unit 1 was in Mode 4 (hot shutdown) at ambient temperature and at 0 percent of rated thermal power coming out of the lR1O refueling outage. Other than that described herein, there was no inoperable equipment that contributed to the occurrence of this event.

C. DESCRIPTION OF EVENT

On July 28, 2003, an evaluation was completed that determined a safety system functional failure could have occurred following the lRlO refueling outage on April 6 & 7, 2002. Specifically, it was determined that the containment building held debris in an amount adequate to block the containment sump screens, which could result in inadequate net positive suction head and possible failure of the residual heat removal (RHR) pumps to perform at design limiting conditions. This may have occurred following a design basis accident had these pumps been called on to operate and draw suction from the containment sumps. The condition existed for a period of 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> and 12 minutes while the unit was in Mode 4 (hot shutdown) and Mode 3 (hot standby), until the loose debris was removed. A walkdown to verify containment cleanliness removed the miscellaneous loose debris which reduced the cross-sectional area of sump screen blockage to an acceptable level.

At the end of the Unit 2 refueling outage in November 2002, an engineer noticed unjacketed torn insulation at the bottom of the steam generators and raised the prospect that, following a design basis accident, some of the insulation could travel to the containment sump and lead to sump screen blockage. An investigation by design engineering found that additional insulation should be considered destroyed due to the missing jacketing which increases the amount of debris that could travel to the sump screen following the design basis accident. When taking into account the additional sump screen blockage that would be incurred by the insulation, along with other containment debris, design engineering determined that adequate net positive suction head still existed in November 2002 for both the containment spray pumps and the residual heat removal pumps. Therefore, no reportable condition existed.

NRC Form3 J56A (1-=01)U.S. NUCLEAR REGULATORY COMMISSION (1.2001)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE (3)

YEAR I SEQUENTIAL REMISON

.l

.YEAR NUMBER Vogtle Electric Generating Plant - Unit 1 05000-424 2003 -- 001

-- 00 3 OF 4 TEXT Ili more space Is required, use additional copies of NRC Fom 366A) ( 17 As a result of reduced NPSH margin, design engineering then began to review previous startups from refueling outages to determine if debris found in containment after each outage, combined with the additional insulation, could have led to sump screen blockage following a design basis accident.

Potential events were identified and an evaluation of these events was performed to determine the impact on NPSH margin. The results of this review found one event in the last three years where sump screen blockage could have been sufficient to cause inadequate net positive suction head for the RHR pumps. This event occurred on April 6 & 7, 2002, as stated previously.

D. CAUSE OF EVENT

The cause of this event was inadequate removal of debris from the Unit 1 containment building prior to Mode 4 entry. Contributing to this were personnel errors by licensee personnel who performed inadequate walkdowns to ensure that all debris had been removed.

Another contributing factor to the occurrence of this event was the inadequate calculation for sump screen blockage allowable limits by the original architect/engineer. The addition of insulation to the allowable sump screen blockage calculations requires that less loose debris be left in containment, raising the standards for containment cleanup requirements.

E. ANALYSIS OF EVENT

At the time of this event, approximately one-half of the subcritical core consisted of new fuel. The remainder had been removed from the previous core and the fission process for more than a month.

Based on these factors, the reactor core possessed a minimal amount of decay heat.

The total time of this event of 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> and 12 minutes represents a narrow window of opportunity for a design basis LOCA to have occurred, severely reducing the probability of such an event.

Finally, an assumption was made that 100% of the insulation material destroyed during a design basis LOCA would migrate to the sump. This assumption is overly conservative as documented in recent NUREGs. Design engineering believes that the insulation debris created during a LOCA at the operating temperature and pressure stated above would be greatly reduced from that of the design basis LOCA. Insulation destruction is based on testing from a jet blast from a water/steam source at 590'F and 1595 psi. Testing information for insulation destruction at the reduced operating temperature and pressure existing during this event is not available. Additionally, much of the insulation debris will not transport to the sump as it will be caught up on grating, equipment, and pocket areas in containment.

NRC Fam 366A(1-2001)

r iIU.S. NUCLEAR REGULATORY COMMISSION (1-2001)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE 13)

TY SEQUENnLAI REVISION

~~~~~~~~~YEAR NUMBER Vogtle Electric Generating Plant - Unit 1 05000-424 2003 -- 001

-- 00 4 OF 4 TEXT (ff more space is required, use additional copies of NRC Form 366A) 117 Based on these considerations, there was no adverse effect on plant safety or on the health and safety of the public as a result of this event.

This event could have resulted in a safety system functional failure.

F. CORRECTIVE ACTIONS

1) Sufficient debris was removed from containment on April 7, 2002, to enable net positive suction head requirements to be met for the RHR pumps.

2) The calculation for sump screen blockage allowable limits has been revised to include the contribution from the steam generator insulation.

3) Cleanliness walkdowns prior to Mode 4 entry continue to be performed. Additionally, in the interval since this event occurred, management has stressed cleaning up following individual jobs in the containment building rather than waiting for a walkdown prior to Mode 4 entry to clean up the entire building. This practice of maintaining cleanliness is believed to be a more effective method of ensuring debris is not left behind.

4) Other means for preventing containment sump blockage are detailed in the Vogtle Electric Generating Plant's Response to NRC Bulletin 2003-01 dated August 7, 2003.

G. ADDITIONAL INFORMATION

1) Failed Components:

None

2) Previous Similar Events:

There have been no previous similar events in the last three years.

3) Energy Industry Identification System Code:

Containment Spray System - BE Residual Heat Removal System - BP Containment Sumps - NH NRC FIn-. 3665A 1-ZO01i