05000311/LER-2006-005, Regarding Automatic Start of Auxiliary Feedwater Pumps in Mode 4
| ML070040094 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/21/2006 |
| From: | Fricker C Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N06-0480 LER 06-005-00 | |
| Download: ML070040094 (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 |
| 3112006005R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 PSEG Nuclear LLC DEC 21 2006 1 OCFR50.73 LR-N06-0480 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-001 LER 311/06-005 Salem Nuclear Generating Station Unit 2 Facility Operating License No. DPR-75 NRC Docket No. 50-311
SUBJECT:
Automatic Start of Auxiliary Feedwater Pumps in Mode 4 This Licensee Event Report, "Automatic Start of Auxiliary Feedwater Pumps in Mode 4" is being submitted pursuant to the requirements of the Code of Federal Regulations 1 OCFR50.73(a)(2)(iv)(A).
The attached LER contains no commitments. Should you have any questions or comments regarding this submittal, please contact Mr. Howard Berrick at 856-339-1862.
Sincerel Car
- J.Frricker Salem Plant Manager Attachments (1) 95-2168 REV. 7/99
DEC81 2006 Page 2 Document Control Desk LR-N06-0480 C
Mr. Samuel Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission Attn: Mr. S. Bailey, Licensing Project Manager - Salem Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the
- 13. PAGE Salem Generating Station - Unit 2 05000311.1 OF 4
- 4. TITLE Automatic Start of Auxiliarv Feedwater Pumps in Mode 4
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MNH DY YA AIIYNM OKTNME NUMBER NO.
MNH DY YA FACILITY NAME DOCKET NUMBER 10 29 2006 2006 005 00 12 21 2006
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[D 50.73(a)(2)(vii) 4 E) 20.2201(d)
El 20.2203(a)(3)(ii)
F-50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
-- 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
_E 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL [E 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
[D 50.73(a)(2)(iv)(A)
Ml 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[I 50.36(c)(2)
[E 50.73(a)(2)(v)(A)
El 73.71(a)(4)
[] 20.2203(a)(2)(iv) 0l 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5) 000 El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[)
50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below nr in NRr.
Fnrrn lERA
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (include Area Code)
Howard G. Berrick, Senior Licensing EngineerL 856-339-1862CAUSE sYS COMPONENT MANU-REPORTABLE I
SYSTEM MANU-REPORTABLE TEM C
FACTURER TO EPIX
CAUSE
COMPONENT FACTURER TO EPIX A
- - _No
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
ED NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On October 29, 2006, with Unit 2 in Mode 5 and starting up from its fifteenth refueling outage, Main Turbine surveillance testing (ST) was in progress with the turbine governor and stop valves open. The ST procedure allows testing with the plant to in Mode 5, Mode 6, or Mode 4 with the main steam isolated from the main turbine. Later in the day, the plant transitioned into Mode 4. This mode change procedurally required that the Main Steam Isolation Valves (MSIVs) and bypass valves be closed; however, not all members of the operating shift realized the requirement.
During the performance of the turbine testing, post maintenance testing activities on the 22 loop MSIV were requested. Some of the shift Operations personnel were briefed on the opening of the MSIV for retest, but failed to recognize that the Main Turbine stop and governor valves were open. As the MSIV stroked opened, the 22 Steam Generator (S/G) narrow range indicated level decreased, generating an Engineered Safety Features automatic start signal for the 21 and 22 motor driven Auxiliary Feedwater pumps. Operators promptly restored indicated feedwater level to 22 S/G. The cause of this event is attributed to inadequate oversight of plant activities and is reportable in accordance with 1 OCFR50.73 (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) of this section."
NRC FORM 366 16-2004)
PRINTED ON RECYCLED PAPER NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPER
(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 Form 366A)
SAFETY CONSEQUENCES AND IMPLICATIONS
There was no actual safety consequences associated with this event; sufficient cooling was always maintained. The low level in the 22 S/G occurred as a result of human error and was not caused by equipment malfunction. The safety systems responded to the low S/G level as designed.
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. There was no condition that alone could have prevented the fulfillment of a safety function of a system needed to remove residual heat.
CORRECTIVE ACTIONS
- 1. All control room activities for the duration of the outage were re-verified to ensure that no operational conflicts existed.
- 2. The pre-job brief used for performance of MSIV testing was updated to include this event to be used prior to future testing.
- 3. A lessons-learned briefing to all licensed operators will be performed prior to the next refueling outage. This will reinforce expectations pertaining to roles and responsibilities of the person having command and control of the unit during periods when multiple activities are being performed.
- 4. A readily available means in the control room to visually identify key operating precautions and limitations in progress during outages, which can be used to recognize potential testing activity conflicts, will be developed.
COMMITMENTS
No commitments are made in this LER,