05000247/LER-2012-004, For Indian Point Unit 2, Regarding Unanalyzed Condition and Safety System Functional Failure Due to Use of Rad Bypass Switch for Steam Generator Blowdown Isolation Valves Which Defeats Their Automatic Isolation for Analyzed Eve
| ML12158A190 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  (DPR-026) |
| Issue date: | 05/25/2012 |
| From: | Ventosa J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-12-057 LER 12-004-00 | |
| Download: ML12158A190 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2) |
| 2472012004R00 - NRC Website | |
text
.Entergy Indian PAint Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 John A. Ventosa Site Vice President NL-12-057 May 25, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2012-004-00, "Unanalyzed Condition and Safety System Functional Failure due to Use of Rad Bypass Switch for Steam Generator Blowdown Isolation Valves which Defeats Their Automatic Isolation for Analyzed Events" Indian Point Unit No. 2 Docket No. 50-247 DPR-26
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2012-004-00. The attached LER identifies an event where there was an unanalyzed condition and a safety system functional failure due to use of the Rad Bypass switch for Steam Generator blowdown isolation valves during testing which defeats their automatic isolation for analyzed events. This condition is reportable under 10 CFR 50.73(a)(2)(ii)(B) and 10CFR50.73(a)(2)(v)(B). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2012-02408.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 254-6710.
Sincerely, JAV/cbr
/W"J.--.4 10AI cc:
Mr. William Dean, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 2 Mrs. Bridget Frymire, New York State Public Service Commission LEREvents@inpo.org
Abstract
On March 27,
- 2012, it was identified that use of the Rad Bypass switch position for Steam Generator (SG) blowdown isolation valves would defeat the automatic isolation of the blowdown isolation valves for degraded heat sink events (Loss of Normal Feedwater, Loss of All AC Power to Station Auxiliaries).
The normal valve position is open and Auto close for heat sink events in addition to containment Phase A isolation.
These analyzed events assume SG blowdown (SGBD) isolation occurs and continuous SG blowdown during these events has not been analyzed.
SG inventory would not be maintained because one Auxiliary Feedwater pump would not provide adequate flow with the blowdown isolation valves open.
On January 19,
- 2011, test 2-PC-2Y23-49 (Liquid Radiation Monitor Calibration) was initiated which positioned all SG blowdown isolation valve switches to Rad Bypass.
During this time on January 20,
- 2011, the 21 Auxiliary Feedwater Pump was removed from service for testing.
On January 27, 2011, testing per 2-PC-2Y23-49 was completed.
The apparent cause was the inappropriate revision of test procedure 2-PC-2Y23-49 in 2002 that deleted information previously incorporated from an operating event at unit 3 that restricted when the blowdown radiation monitor could be tested.
Corrective actions included revision of procedure 2-PC-2Y23-49 to delete steps to place in Rad Bypass while performing Radiation Monitor R-49 calibration and installation of a test jumper to disable the blowdown function.
The UFSAR will be revised to provide the assumptions credited in accident analysis for SGBD isolation.
The event had no significant effect on public health and safety.
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17)
Corrective Actions
The following corrective actions have been performed under the Corrective Action Program (CAP) to address the cause of this event.
" Procedure 2-PC-2Y23-49 was revised to delete steps to place in Rad Bypass while performing Radiation Monitor R-49 calibration and to install a test jumper to disable the blowdown function.
" The UFSAR will be revised to include the assumptions in the applicable accident analysis that SGBD isolation is assumed from event initiation.
Event Analysis
The event is reportable under 10CFR50.73(a) (2) (ii)
(B) and 10CFR50.73(a) (2) (v).
The licensee shall report any condition that resulted in:
(B)
The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety, and any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (B)
Remove residual heat.
This event meets the reporting criteria because placement of the switch for the SG Blowdown isolation valves in Rad Bypass defeated their Auto close function for degraded heat sink events (Loss of Normal Feedwater, Loss of All AC Power to Station Auxiliaries).
Failure to close the SG blowdown isolation valves would result in inadequate maintenance of SG inventory because one motor driven Auxiliary Feedwater pump would not provide adequate flow with the blowdown isolation valves open.
During the test, one motor driven AFWP was inoperable.
These analyzed events assume SG blowdown isolation occurs and continuous SG blowdown during these events has not been analyzed.
The condition was a safety system functional failure (SSFF) since during the test one motor driven AFWP was out of service resulting in inadequate maintenance of SG inventory.
In accordance with reporting guidance in NUREG-1022, for a SSFF per 10CFR50.73(a) (2) (v,)an additional random single failure need not be assumed during the condition.
Past Similar Events A review was performed of the past three years of Licensee Event Reports (LERs) for events reporting control switches that could defeat or bypass automatic design features.
No Unit 2 LERs were identified.
Safety Significance
This event had no significant effect on the health and safety of the public. There were no actual safety consequences for the event because there were no applicable accidents or transients (LONF or LOOP) during the time testing was performed by 2-PC-2Y23-49.
For postulated events, the turbine driven AFWP was available and capable of providing feedwater flow to maintain SG inventory during the time testing was being performed with the SG blowdown isolation valve switches positioned in Rad Bypass.
Administrative controls and instrumentation are available for operators to mitigate this condition.
For the LONF or LOOP events, procedure 2-E-0 (Reactor Trip or Safety Injection) would be entered and transition to 2-ES-0.1 SG inventory will be impacted but plant procedure 2-ES-0.1 will ensure SG narrow range level is reestablished in all SGs to maintain symmetric cooling of the RCS.
This procedure includes a step to verify that the SG blowdown isolation valves are closed.
Additionally, SG blowdown is normally throttled by valve MS-71 to approximately 20 gpm even with the SG blowdown isolation valves open.
During the January 19-27, 2011 testing per 2-PC-2Y23-49, SG blowdown was approximately 25 gpm per SG.