05000311/LER-2010-001, Regarding Manual Reactor Trip Due to Degraded Condenser Heat Removal
| ML100640546 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/03/2010 |
| From: | Fricker C Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N10-0069 LER 10-001-00 | |
| Download: ML100640546 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3112010001R00 - NRC Website | |
text
CO PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC 1 OCFR50.73 MAR 0 3 2010 LR-N 10-0069 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-001 LER 311/2010-001 Salem Nuclear Generating Station Unit 2 Facility Operating License No. DPR-75 NRC Docket No. 50-311
Subject:
Manual Reactor Trip Due to Degraded Condenser Heat Removal This Licensee Event Report, "Manual Reactor Trip Due to Degraded Condenser Heat Removal" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a)(2)(iv)(A).
The attached LER contains no commitments. Should you have any questions or comments regarding this submittal, please contact Mr. E. H. Villar at 856-339-5456.
Since 0 Z jJ. Friicker Site Vice President - Salem Attachments (1) 95-2168 REV. 7/99
Document Control Desk MAR, 03 2010 Page 2 LR-N10-0069 cc Mr. S. Collins, Administrator - Region I Mr. R. Ennis, Licensing Project Manager - Salem USNRC Senior Resident Inspector - Salem (X24)
Mr. P. Mulligan, Manager IV Mr. H. Berrick, Salem Commitment Tracking Coordinator Mr. L. Marabella, Corporate Commitment Tracking Coordinator
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Salem Generating Station - Unit 2 05000311 T
1 of 4
- 4. TITLE Manual Reactor Trip Due to Degraded Condenser Heat Removal
- 5. EVENT DATE
- 6. LER NUMBER__
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEUNILRVFACILITY NAMEDCETNME MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR NUMBER NO.
DOCKET NUMBER 01 03 201020100 0 1 0 03 03 2010
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 0] 20.2201(b) 0l 20.2203(a)(3)(i) 0l 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 0E 20.2201(d)
El 20.2203(a)(3)(ii) 0l 50.73(a)(2)(ii)(A) 0l 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[I 20.2203(a)(4)
[I 50.73(a)(2)(ii)(B) 0l 50.73(a)(2)(viii)(B)
[: 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii) 0l 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 0l 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A)
N 50,73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
[] 20.2203(a)(2)(iii) 0l 50.36(c)(2)
[I 50.73(a)(2)(v)(A) 0l 73.71 (a)(4)
El 20.2203(a)(2)(iv) 0l 50.46(a)(3)(ii)
[I 50.73(a)(2)(v)(B) 0l 73.71(a)(5) 80%
0l 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A) 0l 50.73(a)(2)(v)(C) 0l OTHER [E
20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
[E 50.73(a)(2)(v)(D)
Specify in Abstract below or in DESCRIPTION OF OCCURRENCE (cont'd) 0745 21 B circulator emergency tripped due to high TWS differential pressure.
0747 Commenced reducing main turbine load at 1% per minute to 80% power in accordance with abnormal operating procedures.
0748 23A circulator emergency tripped due to high TWS differential pressure.
0754 21B circulator placed in service and then emergency tripped due to high TWS differential pressure.
0803 21A circulator emergency tripped due to high TWS differential pressure.
0804 Unit 2 reactor is manually tripped from approximately 80% due to having only two circulators (22A and 228) in service. Control room operating personnel transitioned to Emergency Operating Procedure Trip 1. As expected, the Auxiliary Feedwater Pumps started automatically to feed the steam generators.
0837 Unit 2 is stabilized in Mode 3 in accordance with plant operating procedures.
The unit was returned to service on January 4, 2010, at 1831 after ice was cleared from the screens, condenser water boxes were cleaned and demonstrated to operate properly.
This report is being made in accordance with 10CFR50.73(a)(2)(iv)(A), "any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)...."
CAUSE OF OCCURRENCE The cause of the manual reactor trip is attributed to high amounts of river ice and resultant accumulation of ice on the CW traveling water screens. The heavy ice loading caused the high TWS differential pressure. A TWS differential pressure of ten (10) feet causes a circulating water pump trip.
PREVIOUS OCCURRENCES
A review of LERs at Salem Station dating back to 2003 identified two other occasions of a reactor trip due to unusual harsh environmental conditions. LER 311/2003-001 "Manual Reactor Trip Due to Degradation of Condenser Heat Removal" and LER 272/2007-002 "Manual Reactor Trips Due to Degraded Condenser Heat Removal" were caused by excessive grassing. The corrective actions associated with these events were intended to improve the reliability of the CW system in responding to excessive river grass; and therefore would not have been expected to prevent the occurrence of this ice related event.
SAFETY CONSEQUENCES AND IMPLICATIONS
There was no actual safety consequence associated with this event. Operators appropriately responded to the degraded CW system (loss of circulating water pumps) and the potential loss of normal heat sink (condenser) by manually tdpping the reactor in accordance with plant procedures.
Plant response to the manual reactor trip was normal. All safety systems operated as required, including the service water system which also takes suction from the river.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in NEI 99-02, Regulatory Assessment Performance Indicator Guidelines, did not occur. The event did not result in a condition that alone could have prevented the fulfillment of a safety function of a system needed to remove residual heat.
CORRECTIVE ACTIONS
- 1. Circulating water traveling water screens and condenser water boxes were cleaned, inspected and retumed to normal service.
- 2. A root cause evaluation was initiated to identify other potential causes and corrective actions to improve response to severe weather conditions.
- 3. This evaluation and any identified improvements will be tracked in PSEG's corrective action program.
COMMITMENTS
No commitments are made in this LER.