05000280/LER-2003-006, Regarding Steam Generator AFW Isolation Unanalyzed Condition from Original Design
| ML040480591 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/09/2004 |
| From: | Blount R Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 03-636 LER 03-006-00 | |
| Download: ML040480591 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2802003006R00 - NRC Website | |
text
1 OCFR50.73 Virginia Electric And Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 February 9, 2004 U. S. Nuclear Regulatory Commission Serial No.:
03-636 Attention: Document Control Desk SPS: BAG/TJN R' Washington, D. C. 20555-0001 Docket No.:
50-280 50-281 License No.:
DPR-32 DPR-37
Dear Sirs:
Pursuant to 10CFR50.73, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Surry Power Station Units 1 and 2.
Report No. 50-280, 50-281/2003-006-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Management Safety Review Committee for its review.
Very truly yours, Richard H. Blount, Site Vice President Surry Power Station Enclosure Commitment contained in this letter:
- 1. A modification to install stop check valves downstream of each of the auxiliary feedwater steam generator isolation motor operated valves, prior to where the two headers join together, will be performed during the refueling outages on Unit 1 and Unit 2.
cc:
United States Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23 T85 Atlanta, Georgia 30303-8931 Mr. G. J. McCoy NRC Senior Resident Inspector Surry Power Station
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EVENT R P T (
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FACILIlY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
SURRY POWER STATION, Unit 1 l 05000 - 280 1 OF 5 TITE (4)
Steam Generator AFW Isolation Unanalyzed Condition from Original Design EVENT DATE (5) ll LER NUMBER (6)
- - REPORT DATE (7)]l OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY FACILITY NAME DOCKETNUMBER NUMBER NUMBER DAYAR Surry Power Station, Unit 2 05000-281 12 12 2003 2003
-- 006 --
00 2
9 2004 FACLITY NAME DOCKETNUMBER l1____ I !. ii 1 05000-OPERATING l
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 0CFR §: (Check all that apply) (11)
MODE (9) lN 20.2201(b) 20.2203(a)(3)(ii)
X 50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)
POWER 20.2201(d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LEVEL (10) 100 l 20.2203(a)(1) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) l 73.71(a)(4) 20.2203(a)(2)(i) l 50.36(C)(1)(ii)(A) 50.73(a)(2)(v)(A) 73.71 (a)(5)
Irk ZZZII l 20.2203(a)(2)(ii) l 50.36(c)(2)
I50.73(a)(2)(v)(B)
OTHER 20.2203(a)(2)(iii) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C)
Specify In Abstract below or 20.2203(a)(2)(iv) l 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D)
In (If more space Is required, use additional copies of NRC Forn 366A) (17)
On December 12, 2003 at 1119 hours0.013 days <br />0.311 hours <br />0.00185 weeks <br />4.257795e-4 months <br />, with Units 1 and 2 at 100% reactor power, a second plant issue was submitted documenting the results of the review. Unit 1 and Unit 2 AFW systems were declared inoperable and a 6-hour clock to hot shutdown (HSD) was entered for both units in accordance with Technical Specification (TS) 3.0.1.
At 1433 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.452565e-4 months <br />, on December 12, 2003, an eight-hour notification was made to the NRC of the unanalyzed condition pursuant to 10 CFR 50.72(b)(3)(ii)(B).
This report is provided pursuant to 10 CFR 50.73(a)(2)(i)(B), for the inoperable AFW conditions prohibited by TSs, and 10 CFR 50.73(a)(2)(ii)(B), for the plant being in an unanalyzed condition.
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
The existing accident analysis for a SGTR assumes auxiliary feedwater isolation to a ruptured steam generator within 30 minutes. Isolation of AFW to a ruptured SG may not be possible within the time frame specified in the analysis.
For this specific SGTR scenario, failure to isolate AFW could result in a radioactive release or failure of equipment important to safety, which has not been previously analyzed.
A probabilistic risk assessment found that the unisolable steam generator increased the risk contribution from a SGTR sequence, however, the actual risk contribution was small, and was a negligible portion of the total core damage frequency.
For this specific scenario the contribution was less than 6E-9 per year.
Given these considerations, the interim measures and actions discussed in Sections 4.0 and 5.0, and the fact that a SGTR concurrent with loss of one emergency bus did not occur, this condition resulted in no significant safety consequences, and the health and safety of the public were not affected.
3.0 CAUSE
The current AFW configuration has existed since original construction.
No documentation was found that indicated the AFW MOVs have ever been required to be in a special alignment to account for this condition. The root cause for this condition is design configuration/ analysis, in that the unanalyzed condition was not considered during initial plant design.
The previous simulator computer did not model backflow through the AFW MOVs.
A recent upgrade to the model more accurately reflected the plant design and showed that the steam generator in the simulator training session continued to fill after the energized AFW MOVs were closed.
(If more space is required, use additional copies of (f more space is required, use additional copies of NRC Form 366A) (17)
7.0 SIMILAR EVENTS
None.
8.0 MANUFACTURER/MODEL NUMBER Not applicable.
9.0 ADDITIONAL INFORMATION
None.