05000373/LER-2015-002, Regarding Valve Control Power Breaker-Fuse Coordination Issue Results in Unanalyzed Condition
| ML15041A831 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 02/10/2015 |
| From: | Vinyard H Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-15-004 LER 15-002-00 | |
| Download: ML15041A831 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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 10 CFR 50.73(b)(2)(ii)(B) |
| 3732015002R00 - NRC Website | |
text
ion 10 CFR 50.73 RA1 5-004 February 10, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-18 NRC Docket Nos. 50-373 and 50-374
Subject:
Licensee Event Report 2015-002-00 Valve Control Power Breaker-Fuse Coordination Issue Results in Unanalyzed Condition In accordance with 10 CFR 50.73(a)(2)(ii)(B), Exelon Generation Company (EGC), LLC, is submitting Licensee Event Report Number 2015-002-00 for LaSalle County Station Units 1 and 2.
There are no regulatory commitments in this letter. Should you have any questions concerning this report, please contact Mr. Guy V. Ford, Regulatory Assurance Manager, at (815) 415-2800.
Harold T. Vinyard Plant Manager LaSalle County Station
Enclosure:
Licensee Event Report cc:
Regional Administrator - NRC Region III NRC Senior Resident Inspector - LaSalle County Station
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
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 />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections LICENSEE EVENT REPORT (LEIR}
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by J
I I f ll il nternet e-ma to n oco ects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of 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.
- 3. PAGE LaSalle County Station Units 1 and 2 05000373 1
OF 3
- 4. TITLE Valve Control Power Breaker-Fuse Coordination Issue Results in Unanalyzed Condition
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME LaSalle County Station Unit 2 DOCKET NUMBER 05000374 12 12 2014 2015 -
002 00 02 10 2015 FACILITY NAME N/A DOCKET NUMBER N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §:
(Check all that apply) q 20.2201(b) q 20.2203(a)(3)(i) q 50.73(a)(2)(i)(C) q 50.73(a)(2)(vii) q 20.2201(d) q 20.2203(a)(3)(ii) q 50.73(a)(2)(ii)(A) q 50.73(a)(2)(viii)(A) q 20.2203(a)(1) q 20.2203(a)(4) 50.73(a)(2)(ii)(B) q 50.73(a)(2)(viii)(B) q 20.2203(a)(2)(i) q 50.36(c)(1)(i)(A) q 50.73(a)(2)(iii) q 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL q 20.2203(a)(2)(ii) q 50.36(c)(1)(ii)(A) q 50.73(a)(2)(iv)(A) q 50.73(a)(2)(x) q 20.2203(a)(2)(iii) q 50.36(c)(2) q 50.73(a)(2)(v)(A) q 73.71(a)(4) q 20.2203(a)(2)(iv) q 50.46(a)(3)(ii) q 50.73(a)(2)(v)(B) q 73.71(a)(5) 100 q 20.2203(a)(2)(v) q 50.73(a)(2)(i)(A) q 50.73(a)(2)(v)(C) q OTHER q 20.2203(a)(2)(vi) q 50.73(a)(2)(i)(B) q 50 73(a)(2)(V)(D) fy in ct below or in NRC F orm 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT TELEPHONE NUMBER (Include Area Code)
Richard Meyer, Design Engineering Manager Electrical 815-415-2514CAUSE SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX
CAUSE
SYSTEM COMPONENT MANU FACTURER REPORTABLE TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR F1 YES I'll "yes, complete 15. EXPECTED SUBMISSION DATE)
NO SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On December 12, 2014, both Units 1 and 2 were in Mode 1 at 100% power with an NRC Triennial Fire Protection Inspection in progress. During the inspection, breaker-fuse coordination evaluations were conducted for the 250 VDC control power supplies to Reactor Core Isolation Cooling (RCIC) valves that are required to be operated from the Remote Shutdown Panel (RSP) in the event of a fire in the Main Control Room (MCR). The evaluation identified that the power supply breakers to two RCIC valves might trip before the fuses, requiring the breakers to be reset before they could be operated from the RSP. The plant current licensing basis does not identify the contingency actions required to reset the breakers prior to operation from the RSP. RCIC operation from the RSP is the method credited for reactor vessel inventory makeup in the event of a fire in the MCR. This condition is common to both Units.
Documented breaker-fuse analyses were not required by engineering standards in place during original construction.
Corrective actions were to implement compensatory measures as identified in Information Notice 97-48, "Inadequate or Inappropriate Interim Fire Protection Compensatory Measures," including communicating the issue to the operating crews and providing direction to the Safe Shutdown Equipment Operator in the event of a fire in the MCR. Measures to reset the breakers will be added to the appropriate procedures until the breakers and/or trip settings are appropriately modified.
NRC FORM 366 (02-2014)
LaSalle County Station Units 1 and 2 are General Electric Company Boiling Water Reactors with 3546 Megawatts Rated Core Thermal Power.
A.
CONDITION PRIOR TO EVENT
Unit(s): 1 /2 Event Date: December 12, 2014 Event Time: 1500 CST Reactor Mode(s): 1/1 Mode(s) Name: Power Operation Power Level: 100%/100%
B.
DESCRIPTION OF EVENT
On December 12, 2014, both Units 1 and 2 were in Mode 1 at 100% power with an NRC Triennial Fire Protection Inspection in progress. During the inspection, it was identified that there were no documented breaker-fuse coordination analyses for the 250 VDC control power supplies to Reactor Core Isolation Cooling (RCIC)[BN] valves that are required to be operated from the Remote Shutdown Panel (RSP)[JL] in the event of a fire in the Main Control Room (MCR). Breaker-fuse coordination evaluations were conducted, and found that the power supply breakers to two RCIC valves might trip before the fuses, requiring the breakers to be reset before they could be operated from the RSP. The plant current licensing basis does not identify the contingency actions required to reset the breakers prior to operation from the RSP. RCIC operation from the RSP is the method credited for reactor vessel inventory makeup in the event of a fire in the MCR. This condition is common to both Units.
The affected valves are 1(2) E51-F031, the RCIC suppression pool suction valves, and 1(2) E51 -F046, which supply lubricating oil cooling for the RCIC turbine. If the control power supply breakers for these valves tripped, they could not be operated from the RSP without locally resetting the breakers.
This occurrence is reportable under 10 CFR 50.73(b)(2)(ii)(B) as an event or condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety. An ENS report was made to the NRC (EN 50675) at 2115 ET on December 12, 2014, pursuant to 10 CFR 50.72(b)(3)(ii)(B).
C.
CAUSE OF EVENT
This condition has existed since original construction. Documented breaker-fuse analyses were not required by engineering standards in place at that time.
D.
SAFETY ANALYSIS
The safety significance of this event was minimal. The Technical Specification 3.5.3 operability of RCIC was not affected by this condition.
In the event of a fire in the MCR, RCIC would be available from the RSP to maintain vessel inventory for sufficient time for the valve failure(s) to be identified and diagnosed, and for operators to reset the breakers and open the valves. The normally closed 1(2) E51-F031 valve is not required to establish suction from the suppression pool until low water level in the Condensate Storage Tank is reached, which takes approximately
four hours based on station blackout coping studies. Additionally, evaluation has shown that the RCIC turbine can run for approximately three hours without lube oil cooling in the event that the 1(2)E51-046 valve does not open.
E.
CORRECTIVE ACTIONS
Information Notice 97-48, "Inadequate or Inappropriate Interim Fire Protection Compensatory Measures," was reviewed and compensatory measures implemented, including the issuance of a Standing Order communicating the issue to the operating crews and directing the Safe Shutdown Equipment Operator to report to the respective 250 VDC switchgear to reset any tripped breakers in the event of a fire in the MCR.
Contingency measures to reset the breakers as needed in the event of a fire in the MCR once control has been transferred to the RSP will be added to the appropriate procedures.
The breakers and/or trip settings will be modified for the affected RCIC valves as required.
F.
PREVIOUS OCCURRENCES
A search of LaSalle LERs over the last 10 years did not identify any previous events related to breaker-fuse coordination issues.
G. COMPONENT FAILURE DATA
No component failures occurred during this event.