ENS 46707
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ENS Event | |
|---|---|
01:00 Mar 30, 2011 | |
| Title | Potential Voiding in Auxiliary Feedwater Alternate Suction Line |
| Event Description | The design of the Auxiliary Feedwater (AF) system is for the AF pumps to normally take suction from the condensate storage tank. If the condensate storage tank is not available, the essential service water system provides the alternate supply. Due to the AF system suction piping and valve configuration, a voided section of pipe could exist in the portion that isolates the condensate storage tank supply from the essential service water supply. A preliminary vendor analysis has determined that the void fraction to reach the pump in a dynamic scenario exceeds the acceptance criteria for AF pump operability. Based on past operation in this configuration, the event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). Further review of the void model and pump performance characteristics are planned.
In 2011, prior to the completion of this analysis, the void was refilled and verified full for the 'B' trains at Braidwood U1 and U2. Filling the voided piping of both 'A' trains at Braidwood U1 and U2 is in progress. Once filled, the AF systems are operable. The licensee has notified the NRC Resident Inspector. |
| Where | |
|---|---|
Braidwood  Illinois (NRC Region 3) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | |
| Time - Person (Reporting Time:+0.57 h0.0238 days <br />0.00339 weeks <br />7.80786e-4 months <br />) | |
| Opened: | Joe Klevorn 01:34 Mar 30, 2011 |
| NRC Officer: | Howie Crouch |
| Last Updated: | Mar 30, 2011 |
| 46707 - NRC Website
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Unit 1 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (100 %) |
| After | Power Operation (100 %) |
Unit 2 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (100 %) |
| After | Power Operation (100 %) |
Braidwood with 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat, 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | |
WEEKMONTHYEARENS 533532018-04-20T15:42:00020 April 2018 15:42:00
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