05000272/LER-2007-002, Manual Reactor Trips Due to Degraded Condenser Heat Removal
| ML071930263 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/25/2007 |
| From: | George Gellrich Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N07-0157 LER 07-002-00 | |
| Download: ML071930263 (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 |
| 2722007002R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC Jfl l2 5 2007 1 OCFR50.73 LR-N07-0157 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-001 LER 272/07-002 Salem Nuclear Generating Station Unit 1 Facility Operating License No. DPR-70 NRC Docket No. 50-272 SUBJECT: Manual Reactor Trips Due to Degraded Condenser Heat Removal This Licensee Event Report, "Manual Reactor Trips 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.
Sincerely, George H. Gellrich Salem - Plant Manager Attachments (1) 95-2168 REV. 7/99
Document Control Desk Page 2 JUN 2 5 2007 LR-N07- 0157 cc Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)
Mr. P. Mulligan, Manager IV Bureau of Nuclear Engineering PO 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
- 3. PAGE Salem Generating Station - Unit 1 050002721 1 OF 4
- 4. TITLE Manual Reactor Trips Due to Degraded Condenser Heat Removal
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO.
MNH DY YA FACILITY NAME DOCKET NUMBER 04 24 2007 2007 002 00 06 25 2007
- 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)
El 50.73(a)(2)(vii)
[1 20.2201(d)
El 20.2203(a)(3)(ii)
Ml 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[E 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 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 El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
Z 50.73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
[I 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4)
[E 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5) 40%
[E 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[E 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 NPr. F-nrm IRRA
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME 1TELEPHONE NUMBER (Include Area Code)
Enrique Villar, Senior Licensing Engineer
ý 856-339-5456CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TO EPIX C
KE N
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION [I 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 April 24, 2007 at approximately 2248, a manual reactor trip was initiated with reactor power level at approximately 40%.
The manual reactor trip was initiated in response to a degraded Circulating Water System (CWS) and in accordance with operating procedures. The degradation of the CWS was due to extremely heavy river debris loadings that affected the ability of plant equipment to operate under the condition.
The unit was returned to service on April 26, 2007 at 1236 following the cleaning of the condenser water boxes and lowering river debris conditions.On April 30, 2007, at approximately 1502, with the unit at approximately 80% power a sudden localized high amount of river water debris again affected the CWS intake resulting in the loss of four (4) circulating water pumps due to high screen differential levels, and a manual reactor trip.
The cause of manual reactor trips is attributed to unusual and severe external environmental conditions resulting in record high amounts of river debris loadings, which challenged the design of the CWS. Following cleaning and inspection of the Circulating Water Traveling Screens, condenser waterboxes, and lowering river debris conditions the unit was returned to service on May 3, 2007 at 0310. A root cause investigation has been initiated to improve the plant operation and to minimize challenges to the CWS design by: (a) Improving predictive tools and actions on high river debris conditions, and (b)
Improving the availability and reliability of the Circulating Water System.
This report is being made 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)...."
(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 were no actual safety consequences associated with this event. Operators appropriately responded to the degraded CWS (loss of circulating water pumps) and the potential loss of normal heat sink (condenser) by manually tripping the reactor. Plant response to the manual reactor trips was normal. All safety systems operated as required.
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. Circulating Water traveling screens and condenser waterboxes were cleaned, inspected and returned to service.
- 2. A root cause evaluation was initiated to identify causes and initiate corrective actions that can be made to:
a) Improve predictive tools and actions on high river debris conditions, and b) Further improve the availability and reliability of the Circulating Water System.
COMMITMENTS
No commitments are made in this LER.