05000529/LER-2003-004, Mode 3 Entry with an Auxiliary Feed Water Pump Inoperable - Technical Specification Violation
| ML041630152 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/27/2004 |
| From: | Danni Smith Arizona Public Service Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 192-01141 -DMS/SAB/REB LER 03-004-00 | |
| Download: ML041630152 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) |
| 5292003004R00 - NRC Website | |
text
LAPS A subsidiary of Pinnacke Ne.st Capital Corporation Palo Verde Nuclear Generating Station 1 OCFR50.73 David M. Smith Plant Manager Nuclear Production Tel.
623-393-6116 Mail Station 7602 Fax.
623-393-6077 RO. Box 52034 e-mail: DSMITH10@apsc.com Phoenix, AZ 85072-2034 192-01141 -DMS/SAB/REB May 27, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Unit 2 Docket No. STN 50-529 License No. NPF-51 Licensee Event Report 2003-004-00 Attached please find Licensee Event Report (LER) 50-529/2003-004-00 that has been prepared and submitted pursuant to 1 OCFR50.73. This LER reports a condition prohibited by Technical Specifications in which Unit 2 entered Mode 3 operations with one Auxiliary Feed Water pump inoperable.
In accordance with 1 OCFR50.4, a copy of this LER is being forwarded to the NRC Regional Office, NRC Region IV and the Resident Inspector. If you have questions regarding this submittal, please contact Daniel G. Marks, Section Leader, Regulatory Affairs, at (623) 393-6492.
Arizona Public Service Company makes no commitments in this letter.
Sincerely, DMS/SAB/REB/kg Attachment cc:
B. S. Mallett M. B. Fields N. L. Salgado NRC Region IV Regional Administrator NRC NRR Project Manager + (send electronic and paper)
NRC Senior Resident Inspector for PVNGS
APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004
Abstract
On December 8, 2003 at approximately 0540 MST, Unit 2 was in Mode 4, Hot Shutdown, with heat-up activities in progress to return the unit to service following a refueling outage. A surveillance test was performed to verify that the A train auxiliary feedwater pump discharge check valve was closed. The acceptance criterion of the test was not met and the valve was declared inoperable. Subsequent review by engineering personnel concluded that the test was not valid since the check valve had not been opened with pump flow and the test was not required to be performed until the plant achieved normal operating temperature and pressure. The check valve was then declared operable. The unit entered Mode 3, Hot Standby, and the valve was subsequently successfully tested at NOT/NOP on December 11, 2004.
On April 1, 2004, during an integrated inspection exit meeting, the resident NRC inspector stated that the NRC had concluded that a technical specification violation had occurred when the unit entered Mode 3 since there was insufficient evidence to conclude that the discharge check valve was operable at the time of Mode 3 entry.
The cause of the condition was human error by Engineering and Operations personnel.
NRC FORM 366 Q7-)
(If more space is required, use additional copies of (If more space is required, use additional copies of (If more space Is required, use additional copies of NRC Forn 366A)
December 11 at 0345 MST thus meeting the acceptance criteria of less than or equal to 10 gpm.
On April 1, 2004, during an integrated inspection exit meeting, the resident NRC inspector stated that the NRC had concluded that a technical specification violation of LCO 3.0.4 had occurred when the unit entered Mode 3 on December 8 and again on December 10, since there was insufficient evidence to conclude that the discharge check valve was operable at the time of both Mode 3 entries.
- 5.
ASSESSMENT OF SAFETY CONSEQUENCES
The safety significance of the inoperable discharge check valve is low. The design AF system flow rate is 750 gpm delivered to the intact steam generator(s). The safety analysis flow rate for the AF system is 650 gpm delivered to the intact steam generator(s).
An assessment of the impact of the condition described in this LER (check flow back leakage) was performed by conservatively estimating the affect the condition would have had on the actual flow that would be delivered to the steam generators. The leakage measured during the ST was used to estimate the diversion flow that would be experienced during anticipated accident conditions. The analysis, documented in calculation 02-MA-AF-0042, conservatively estimated the diversion flow through the leaking check valve at 158 gpm (Note: This calculation does not change the design basis for the AF system. It is only for the purpose of evaluating this event). When the allowance for the check valve (158 gpm) is computed, the delivered flow to the steam generators is 544 gpm. Although this value is less than the safety analysis required flow, the value is in excess of the 500 gpm delivered flow to the steam generator(s) which is the success criteria used in the PVNGS PRA. The 500 gpm conservatively assumes reactor core conditions that maximize residual heat. The events described in this report occurred and were resolved prior to unit criticality following unit shutdown for steam generator replacement and refueling. In these circumstances, potential residual heat and, therefore, AF flow requirements would be substantially lower.
There was no safety system functional failure as defined by 10 CFR 50.73(a)(2)(V).
(If more space Is required, use additional copies of (If more space Is required, use additional copies of NRC Form 366A)
- 9.
ADDITIONAL INFORMATION
None