05000311/LER-2014-002, Manual Reactor Trip Due to a Partially Dropped Rod
| ML14090A274 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/31/2014 |
| From: | Jamila Perry Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N14-0081 LER 14-002-00 | |
| Download: ML14090A274 (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 |
| 3112014002R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 PSEG Nuclear LLC LR-N14-0081 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
LER 311/2014-002-00 Salem Nuclear Generating Station Unit 2 Renewed Facility Operating License No. DPR-75 NRC Docket No. 50-311 Manual Reactor Trip Due to a Partially Dropped Rod The Licensee Event Report, "Manual Reactor Trip Due to a Partially Dropped Rod," is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A), as an "... event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)...
The attached LER contains no commitments. Should you have any questions or comments regarding the submittal, please contact David Lafleur of Salem Regulatory Assurance at 856-339-1754.
Sincerely, dW6:
'-*
John F. Perry Site Vice President-Salem Attachments ( 1 )
Document Control Des k Page 2 10 CFR 50.73 LR-N14-0081 cc Mr. W. Dean, Adminis trator-Region 1, NRC Mr. John Hughey, Licensing Project Manager-Salem, NRC Mr. P. Finney, USNRG-Senior Resident Inspector, Salem (X24)
Mr. P. Mulligan, Manager IV, NJBNE Mr. T. Joyce, President and Chief Nuclear Officer-Nuclear Mr. T. Cachaza, Salem Commitment Tracking Coordinator Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. D. Lafleur, Salem Regulatory As surance
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0113112017 (01-2014)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Q.....
Reported lessons learned are incorporated into the licensing process and fed back to industry.
LICENSEE EVENT REPORT {LER)
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by (See Page 2 for required number of internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC digits/characters for each block) 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 13. PAGE Salem Generating Station - Unit 2 05000311 1 OF 4
- 4. TITLE Manual Reactor Trip Due to a Partially Dropped Rod
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR I SEQUENTIAL I REV MONTH FACILITY NAME DOCKET NUMBER DAY YEAR NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 01 31 2014 2014
- - 002
- - 00 03 31 2014 05000
- 9. OEPRATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 1 D 2o.2201(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(i)(C)
D 50.73(a)(2)(vii)
D 20.2201(d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
- 10. POWER LEVEL D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
H 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x) 100%
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4)
D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(C) 0 OTHER D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(v)(D)
Specify in Abstract below or in load adjustments. The control room supervisor directed Reactor Engineering to perform the SDM calculation in accordance with TS 3. 1. 1. 1.
At 0848, operators entered TSAS 3. 2. 1, Action a. for Axial Flux Difference (AFD) outside the target band. TS 3. 2. 1, Action a. 2. a) allows power to continue when operating between 50 and 90 percent RTP provided that the indicated AF D has not been outside the target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and AFD is within limits.
At 0856, reactor power was reduced to less than 75 percent RTP.
At 0904, SDM was determined to be less than 1. 3 percent delta k/k. TSAS 3. 1. 1. 1 requires an immediate initiation of boration at equal to or greater than 33 gpm until required SDM is restored.
Reactor Engineering informed Operations that an RCS boron concentration of 468 ppm would be required to meet the SDM requirement. At 0911, Operators commenced boration at a rate of 33 gpm.
At 0936, reactor power was decreased to less than 50 percent RTP. Operators continued to reduce load for temperature control. TS 3. 1. 1. 1 required SDM had not yet been verified and bqration continued.
At 1001, at approximately 20 percent RTP, as Tave approached 543 degrees Fahrenheit, the reactor was manually tripped. All control rods including rod 102 inserted on the reactor trip. All systems functioned as expected.
At 1004, boration was discontinued. At 1012, Chemistry reported an RCS boron concentration of 520 ppm. TS 3. 1. 1. 1 was exited as SDM requirements were met.
At 1037, operators transitioned from emergency to normal operating procedures, maintaining Operational Mode 3, Hot Standby conditions.
At 1152, a 4 hr. notification was made to the NRC in accordance with the requirements of 10 CFR 50. 72(b )(2)(iv)(B) for an unplanned manual reactor trip.
CAUSE OF OCCURRENCE The cause of dropped rod 1 02 was attributed to a short to ground located in the rod 1 02 Movable Gripper Cable. Two blown fuses in the rod control solid state power cabinet were identified. Subsequent Electrical circuit testing of associated cables indicated a short to ground on the 600 volt power cable from the rod control containment cabinet to the reactor head.
The manual reactor trip was performed due to a lack of available turbine load to continue to maintain Tave on program. With turbine load at 20 percent, operators would have been challenged to continue to maintain Tave on program using turbine load due to the continuous, rapid boration of 33 gpm needed to achieve required SDM with increasing xenon, and end of core life conditions.
PREVIOUS SIMILAR OCCURRENCES A review of LERs at Salem Station dating back to 2011 identified no similar manual reactor trip events.
SAFETY CONSEQUENCES AND IMPLICATIONS
There were no safety consequences associated with this event. Operators appropriately responded to plant conditions to manually trip the reactor and shutdown the plant. All plant safety systems operated as required.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, did not occur.
CORRECTIVE ACTIONS
- 1.
The defective 1 D2 Movable Gripper power cable and rod control solid state power cabinet fuses were replaced. Rod control surveillance testing was performed and rod 102 was returned to service.
- 2.
Additional corrective actions will be determined based on the cause of the rod 1 D2 cable failure and possible enhancements to the SDM calculation procedure.
COMMITMENTS
No commitments are made in this LER.