05000374/LER-2017-003, Regarding High Pressure Core Spray System Inoperable Due to Injection Valve Stem-Disc Separation
| ML17102B424 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/12/2017 |
| From: | Trafton W Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA17-039 LER 17-003-00 | |
| Download: ML17102B424 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v), Loss of Safety Function 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)(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)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3742017003R00 - NRC Website | |
text
Exelon Generation (.
RA17-039 April 12, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit 2 Renewed Facility Operating License No. NPF-18 NRG Docket No. 50-374 LaSalle County Station 2601 North 21 51 Road Marseilles, IL 61341 815-415-2000 Telephone www.exeloncorp.com 10 CFR 50.73
Subject:
Licensee Event Report 2017-003-00, High Pressure Core Spray System Inoperable due to Injection Valve Stem-Disc Separation In accordance with 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(D), Exelon Generation Company, LLC (EGC) is submitting Licensee Event Report (LEA)
Number 2017-003-00 for LaSalle County Station, Unit 2.
There are no regulatory commitments in this letter. Should you have any questions concerning this report, please contact Mr. Guy V. Ford, Jr., Regulatory Assurance Manager, at (815) 415-2800.
Respectfully,
- !ufl!::f!JL Site Vice President LaSalle County Station
Enclosure:
Licensee Event Report cc:
Regional Administrator - NRC Region Ill NRC Senior Resident Inspector - LaSalle County Station
NRCFORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04*2017)
, !he http://www.nri;;.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3D NRC may not conduct or sponsor, and a person is not required to respond lo, the information collection.
- 3. PAGE LaSalle County Station, Unit 2 05000374 1 OF 4
- 4. TITLE High Pressure Core Spray System Inoperable due to Injection Valve Stem-Disc Separation
- 5. EVENT DATE
- 6. LEA NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED I
SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR NA NA FACILITY NAME DOCKET NUMBER 02 11 2017 2017 - 003
- - 00 04 12 2017 NA NA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: {Check all that apply)
D 20.2201 (bl D 20.2203(a)(3)(il D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201 (dl D 20.2203(a)(3)(iil D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B) 5 D 20.2203(a)(1 l D 20.2203(a)(4l D so.13(a)(2)(iiil D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(il D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2l(iil D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 13.11(a)(4l D 20.2203(a)(2)(m)
D so.3s(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71 (a)(S)
D 20.2203(a)(2)(iv)
D so.4s(a)(3)(iil D 50.73(a)(2)(v)(C)
D 13.11(a)(1) 000 D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A) 181 50.73(a)(2)(v)(D)
D 13.11(a)(2)(il D 20.2203(a)(2)(vi) 181 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 13.11(a)(2)(ill D 50.73(a)(2)(i)(C)
D OTHER Specity In Abstract below or In SEQUENTIAL NUMBER 003 REV NO.
00 potential wedge pin failure. At the time of the failure, the station was following vendor and industry guidance for inspection of the valve related to the Flowserve Part 21 Notification.
The causal evaluation into the valve failure is ongoing. The results will be reported in a supplement to this LEA upon completion.
REPORTABILITY AND SAFETY ANALYSIS
This component failure is reportable in accordance with 1 O CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS due to the failure to complete TS 3.5.1, "ECCS - Operating," Required Action 8.2, to restore HPCS system to operable status within the specified Completion Time of 14 days. This component failure is also reportable in accordance with 1 O CFR 50. 73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. This condition could have prevented the HPCS system, a single train safety system, from performing its design function.
There were minimal safety consequences associated with the condition since HPCS was not required to be operable at the time of discovery of the condition, and other required emergency safety systems remained operable. The HPCS injection isolation valve 2E22-F004 was cycled open and closed satisfactorily during previous flow surveillances. There were no actual demands for Unit 2 HPCS, other ECCS systems, or the reactor core isolation cooling (RCIC) system during this period.
This condition is under evaluation as a safety system functional failure defined in accordance with NEI 99-02, Regulatory Assessment Performance Indicator Guideline.
CORRECTIVE ACTIONS
Following discovery of the failure, the Unit 2 HPCS injection isolation valve 2E22-F004 was overhauled using a new stem, and the upper wedge threads were repaired. In addition, a stem lube and rotation check was performed satisfactorily. A review of the diagnostic testing results and pin analysis criteria was performed on all the affected valves in this population. The remaining valves in the population were within the latest vendor's recommendation for rotation criteria.
PREVIOUS OCCURRENCES
A review of station Licensee Event Reports for the past three years, related to stem-disc separation issues, identified the following similar instances:
LEA 374-2014-001: On August 5, 2014, LaSalle County Station Unit 2 automatically scrammed from 100 perc.ent power on high neutron flux, followed by a Group I containment isolation. Following the Group I isolation, the control room operators noted that the position indication for valve 2B21-F022C, the inboard 2C Main Steam Isolation Valve (MSIV), showed dual indication rather than full closed. Troubleshooting of the 2C MSIV determined that the valve stem disc had separated from the stem, which allowed the main disk to drop into the main steam flow path. The resulting reactor pressure transient added positive reactivity, which caused the high neutron flux scram. Increased steam flow in the other three main steam lines resulted in a nearly simultaneous high main steam line flow Group I containment isolation. The cause of the stem-disc separation on the 2C MSIV was fretting wear attributable to marginal design. The root cause of the event was a legacy decision made in 2008 deferring installation of a manufacturer upgrade that would have prevented the failure. Corrective actions include installing the upgrade on all MSIVs on both units, and reviewing previous deferral decisions made using the same decision-making process.
LEA 374-2017-001: On January 23, 2017, operators initiated a manual scram of the LaSalle County Station Unit 2 reactor as a result of observing a generator run-back due to a generator stator winding cooling (GC) system malfunction. Initial troubleshooting identified the most likely cause was plugging in the 'A' GC heat exchanger, based on inspection of the GC system flow-path components. The GC system was realigned to the 'B' heat exchanger until inspections could be performed in the upcoming refueling outage, and the unit was re-started on January 24, 2017. Further inspections of the GC components were performed while the unit was shut down for a planned refueling outage. These inspections determined the cause of the GC system failure was stem-disc separation in the 'A' GC heat exchanger inlet valve. The valve was repaired during the refueling outage. This event is reportable in accordance with 1 O CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation SEQUENTIAL NUMBER 003 REV NO.
00 of the Reactor Protection System (RPS). There were no safety consequences associated with the event since there was no loss of safety function, and the RPS functioned as designed.
LER 374-2017-002: On January 30, 2017, during routine surveillance testing of the LaSalle County Station Unit 2 Division 3 Diesel Generator Cooling Water (DGCW) system, the cooling water strainer backwash valve was unable to open due to stem-disc separation. The valve was replaced, and the HPCS system was returned to operable on February 2, 2017. This condition could have prevented the HPCS system, a single train safety system, from performing its design function. There was minimal safety consequences associated with the event since the other emergency safety systems remained operable, and the Division 3 DGCW system remained functional as it retained the ability to provide the required flow through the system. The apparent cause of the stem-disc separation was erosion due to the carbon-steel valve internals in a raw water system environment.
COMPONENT FAILURE DATA
Manufacturer: Anchor Darling (A391)
Device: Gate Valve, 12-inch Component ID: Model C900 Page _4_ of _4_