05000424/LER-2003-001
| R1) � Docket � Ler � Number 16) | |
| Event date: | |
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| Report date: | |
| 4242003001R00 - NRC Website | |
A. REQUIREMENT FOR REPORT
This event is reportable per 10 CFR 50.73 (aX2Xv)(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)(iXB) 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 I was in Mode 4 (hot shutdown) at ambient temperature and at 0 percent of rated thermal power coming out of the 1R10 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 1R10 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 bead still existed in November 2002 for both the containment spray pumps and the residual heat removal pumps. Therefore, no reportable condition existed.
}IRC Form MBA (1.2001) U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME 11)
DOCKET
05000-424 LER NUMBER (6) I � I MA � IALIMER PAGE 13) 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 RM. 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.
U.S. NUCLEAR REGULATORY COMMISSION DOCKET LER NUMBER 16) � PAGE 131 FACILITY NAME (1) 05000-424 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 vvalkdowns 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