05000269/LER-2009-002-01, Main Steam Relief Valves Exceeded Lift Setpoint Acceptance Band
| ML092800357 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/01/2009 |
| From: | Baxter D Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 09-002-01 | |
| Download: ML092800357 (11) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) |
| 2692009002R01 - NRC Website | |
text
Duke DAVE BAXTER Vice President Oconee Nuclear Station Duke Energy ONO1 VP / 7800 Rochester Highway Seneca, SC 29672 864-873-4460 864-873-4208 fax dabaxter@dukeenergy. com October 1, 2009 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555
Subject:
Oconee Nuclear Station Docket Nos. 50-269,-270,
- - 287 Licensee Event Report 270/2008-02, Revision 1 Problem Investigation Process No.: 0-08-6525, 0-08-7831, 0-06-6400, 0-07-2168, 0-07-5911 Gentlemen:
Pursuant to 10 CFR 50.73 Sections (a) (1) and (d),
attached is Licensee Event Report 270/2008-02, Revision 1, regarding operation with several Main Steam Relief Valves slightly out of tolerance. The report also addresses three prior events which were similar, but not previously recognized as reportable.
This report is being submitted in accordance with 10 CFR 50.73 (a) (2) (i)
(B)
"Any operation or condition prohibited by the plant's Technical Specifications."
Revision 1 corrects a misstatement of fact in the safety analysis section.
Revision 0 stated that only one valve exceeded the 2% criterion used in Duke Energy's safety analysis calculations.
In fact, two valves exceeded this criterion, as indicated by as-found data provided in Revision 0.
It is noted that the two valves were on different Oconee Units.
Therefore, the error does not change our conclusion that this event is considered to be of no significance with respect to the health and safety of the public.
Revision 1 also corrects an unrelated typographical error in a date in the Abstract, standardizes acronyms, and updates the conclusions of the cause investigation.
www. duke-energy. corn
Document Control Desk Date: October 1, 2009 Page 2 There are no regulatory commitments contained in this report.
Any questions regarding the content of this report should be directed to R.P. Todd at 864-873-3418.
Very truly yours, I/
Dave Baxter, Vice President Oconee Nuclear Site Attachment cc:
Mr. Luis Reyes Administrator, Region II U.S. Nuclear Regulatory Commission 61 Forsyth Street, S.
W.,
Suite 23T85 Atlanta, GA 30303 Mr.
John Stang Project Manager U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C.
20555 Mr. Andrew Sabisch NRC Senior Resident Inspector Oconee Nuclear Station INPO (Word File via E-mail)
Abstract
When tested on 10/24/2008, while in Mode 1 prior to shutdown for refueling, as-found lift pressure tests of Oconee Unit 2 main steam relief Valves (MSRVS) revealed 3 unsatisfactory MSRVs out of a total of 16. Technical Specification (TS) 3.7.1 requires 16 MSRVs (8 on each header) to be operable in modes 1, 2 and 3 so Condition A was entered. The affected MSRVs were adjusted, satisfactorily retested, and the condition exited.
Since multiple failures indicate the condition may have arisen over time, there is a
likelihood that all of the required MSRVs were not operable during past plant operation.
Therefore, this occurrence is considered reportable in accordance with 10 CFR 50.73(a) (2) (i)
(B) as a condition prohibited by TS.
A review of prior similar events found three additional occurrences which were not previously recognized as reportable and are addressed in this report (Unit 1, 2 MSRVs, 2006; Unit 2, 2 MSRVs, 2007; and Unit 3, 2 MSRVs, 2007).
The cause of these occurrences has been identified as setpoint drift.
All of the unsatisfactory as-found lift pressures were above the acceptance band but within analysis limits so there was no loss of function.
This event is considered to have no significance with respect to the health and safety of the public.
NRC FORM 366 (7-2001)
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Although the indicated MSRVs failed the owner-established limit of 1%,
they were well within the ASME Code allowed +/-
3% tolerance.
Only two of the valve failures (one on Unit 1 and one on Unit 2) did not remain within the +2 percent tolerance used by the safety analyses that evaluate peak secondary system pressure.
In October of 2006 lMS-8 was found to lift at 1075 psig, or 2.4% above its nominal setpoint.'
In April of 2007 2MS-11 was found to lift at 1104 psig or 2.2% above its nominal setpoint.
The safety analyses that evaluate peak secondary system pressure includes an assumption that one valve (usually the highest setpoint valve) fails to open.
Since the testing demonstrated that the other valves would actually open within values assumed in the safety analyses, having 2MS-11 or lMS-8 open at a slightly higher pressure remains within the bounds of the analyses.
The IMS-8 results were additionally reviewed for any potential impact due to increasing the long-term post-trip steam pressure.
No adverse impact was identified, and lMS-8 remained within the safety analysis assumptions.
As a group, the MSRVs were capable of performing all required safety functions.
Since the valves' actual performance remained within the bounds of the safety analyses, these events had no impact on the predicted results of any accidents and therefore did not. impact the Conditional. Core Damage Probability (CCDP) or Conditional Large Early Release Probability (CLERP).
Therefore, there was no actual impact on the health and safety of the public due to this event.
ADDITIONAL INFORMATION
As stated earlier, a search was performed to identify similar prior events.
This resulted in the additional events documented in this report.
There were no releases of radioactive materials, radiation exposures or personnel injuries associated with this event.
This event is not considered reportable under the Equipment Performance and Information Exchange (EPIX) program.
NRC FORM 366 (7-2001)