05000269/LER-2004-002, Re Main Steam Line Break Mitigation Design/Analysis Deficiency
| ML041950188 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/06/2004 |
| From: | Rosalyn Jones Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 04-002-00 | |
| Download: ML041950188 (10) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2) |
| 2692004002R00 - NRC Website | |
text
Duke R. A. JONES
- wPowere Vice President A Duke Energy Company Duke Power 29672 / Oconee Nuclear Site 7800 Rochester Highway Seneca, SC 29672 864 885 3158 864 885 3564 fax July 6, 2004 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 269/2004-02, Revision 0 Problem Investigation Process No.: 0-04-2808 Gentlemen:
Pursuant to 10 CFR 50.73 Sections (a)(1) and (d),
attached is Licensee Event Report 269/2004-02, Revision 0, regarding a Main Steam Line Break mitigation design/analysis deficiency which could result in the main and startup feedwater control valves being technically inoperable for mitigation of some steam line break scenarios.
This report is being submitted in accordance with 10 CFR 50.73 (a)(2)(i)(B) as a condition prohibited by Technical Specifications, 50.73(a)(2)(ii)(B) as an Unanalyzed Condition, and 50.73(a)(2)(V)(D) as a potential loss of safety function for Accident Mitigation.
This event is considered to be of no significance with respect to the health and safety of the public.
Portions of this report are incomplete.
The root cause investigation and an analysis of the consequences of potentially exceeding the Environment Qualification (EQ) envelope curve are still in progress. At this time we anticipate completing these tasks and providing a supplemented report by approximately August 31, 2004.
- f? § www.duke-energy.com
Document Control Desk Date: July 6, 2004 Page 2 Very truly yours, R. A. Jones Attachment: Licensee Event Report 269/2004-02, Revision 0 cc:
Mr. William D. Travers Administrator, Region II U.S. Nuclear Regulatory Commission 61 Forsyth Street, S. W., Suite 23T85 Atlanta, GA 30303 Mr. L. N. Olshan Project Manager U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C.
20555 Mr. M. C. Shannon NRC Senior Resident Inspector Oconee Nuclear Station INPO (via E-mail)
Date: July 6, 2004 Page 3 bxc: ONS Site:
Document Control (Master File)*
Site PORC Members SA MGR/L.E. Nicholson EPIX Cord/B.R. Loftis Training Mgr/B.K. Jones OPS-Procedures/D.B. Coyle#
RGC: Commitment Index/J.E. Smith#
LER Book*#
Work Control: D.V. Deatherage#
Site Engineering:
W.B. Edge#
K.R. Alter#
GO:
NRIA/M.T. Cash*
ELL/EC050*
NGO/SAA: H.D. Brewer NGO/SA/G.B. Swindlehurst NGO Serv: R.G. Hull#
LEGAL/L.F. Vaughn*
CNS:
SA MGR/R.L. Sweigart MNS:
SA MGR/B.J. Dolan OPS Mgr/S.L. Bradshaw#
PIP FILE*
RGC MGR/B.G. Davenport Vis Cntr Mgr/D.M.Stewart T.A. Ledford#
V.B. Bowman#
M.S. Tuckman#
NSRE/R.L. Gill/EC05N*
C.M. Misenheimer#
RATES/M.J. Brown#
4-RGC MGR/L.A. Keller RGC MGR/C.J. Thomas Non-routine Recipients:
None Hardcopy -
All others by LOTUS NOTES Distribution Copied By Request: -
All others by Directive (Revised 4-5-2004)
Abstract
The Automatic Feedwater Isolation System (AFIS) Circuitry actuates various components including the main and startup feedwater control valves (FCVs) in order to mitigate a Main Steam Line Break (MSLB) with or without a Loss of Offsite Power (LOOP). Tech Spec 3.7.3 requires the FCVs to be operable. The FCVs fail-as-is and require Instrument Air (IA) to close.
On April 29, 2004, Units 1 and 3 were operating in Mode 1 at 100% power; Unit 2 was in No Mode during a refueling outage. During discussion between Site Engineering (SE) and General Office-based Safety Analysis (SA) personnel, it was recognized that, for smaller breaks, actuation signals/alarms may be delayed such that the IA header may depressurize before AFIS and/or operator actions initiate FCV closure. For breaks inside containment, this could lead to pressurization of the Reactor Building (RB) above the RB design pressure (but below RB failure pressure, 144 psig). Immediate action was taken to maintain a diesel air compressor operating at all times pending a more permanent resolution. On May 4, 2004 the event was determined to be reportable. The FCVs are considered to have been inoperable longer than allowed by TS. The root cause investigation is still in progress, this report will be supplemented after it is complete. This event is considered to have no significance with respect to the health and safety of the public.
NRC FORM 366 (7-2DDI)
(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 (if more space Is required, use additional copies of (If more space is required, use additionat copies of (If more space Is required, use additional copies of NRC Form 366A) (17) lE-08.
Even if this entire CDF were conservatively considered to lead to a large early release, the resulting impact would fall well below the risk significant LERF threshold lE-07.
When additional factors are considered such as the specific break size and location, the actual impact is expected to be considerably less.
In particular, the Oconee containment has been shown to be very robust under overpressure conditions (Reference:
Oconee IPE Submittal, Volume III, Appendix G, "Containment Capacity Assessment").
Up to a pressure of approximately 107 psig, the estimated probability of containment failure is less than 1 percent.
The mean containment failure pressure is estimated to be 144 psig.
Any contribution to LERF would be expected to be at least 2 orders of magnitude below the CDF contribution.
Therefore, there was no actual impact on the health and safety of the public due to this event.
ADDITIONAL INFORMATION
The evaluation for recurring problems depends on the root cause classification.
This section will be revised when the report is supplemented.
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.