05000306/LER-1917-003, Regarding Both Containment Spray Pump Control Switches in Pull-Out in Mode 4
| ML18011A041 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/11/2018 |
| From: | Sharp S Northern States Power Company, Minnesota, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-18-004 LER 17-003-00 | |
| Download: ML18011A041 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3061917003R00 - NRC Website | |
text
1717 Wakonade Drive Welch, MN 55089 800.895.4999 xcelenergy.com January 11, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Prairie Island Nuclear Generating Plant Unit 2 Docket 50-306 Renewed License No. DPR-60 Xcel Energy RESPONSIBLE BY NATURE L-PI-18-004 10 CFR 50.73 Licensee Event Report 50-306/2017-003-00, Both Containment Spray Pump Control Switches in Pull-Out in Mode 4 Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), encloses Licensee Event Report (LER) 50-306/2017-003-00, Both Containment Spray Pump Control Switches in Pull-Out in Mode 4.
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
kff~
Scott Sharp Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure cc:
Regional Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC State of Minnesota
Document Control Desk Page 2 ENCLOSURE 1 Licensee Event Report 50-306/2017-003-00 3 Pages Follow
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017)
Eslimated,
httQ://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1 022/r30 the NRC may not conduct or sponsor, and a person Is not required to respond to, the Information collection.
- 13. PAGE Prairie Island Nuclear Generating Plant Unit 2
'05000 306 1 OF3
- 4. TITLE Both Containment Spra Pump Control Switches in Pull-out in Mode 4
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR REV MONTH DAY YEAR NUMBER NO.
05000 11 12 2017 2017
- - 003
- - 00 1
11 2018 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §:(Check all that apply)
Unit 2 Mode 4 D 20.2201(b)
D 20.2203(a)(3)(i) 0 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D 50. 73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i) 0 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)(2)(ii) 0 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
I D 50.73(a)(2)(v)(B)
D 73.71(a)(5)
Unit 2 0%
D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) t8J 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
[8J 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii) 0 50.73(a)(2)(i)(C) 0 OTHER Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT r~LEPHONE NUMBER (Include Area Code)
\\Frank Sienczak
\\(612) 342-8987
- 13. COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE) r;gj NO SUBMISSION DATE
~BSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On November 12, 2017 at 2119, a Control Room board walkdown discovered that both of the Unit 2 Containment Spray Pump control switches had been left in pull-out, when operators transitioned Unit 2 from Mode 5 to Mode 4. With the control switches in pull-out, the pumps would not automatically start as required. Technical Specification (Tech Specs) 3.0.3 was entered as a result of not complying with Technical Specification 3.6.5, Containment Spray and Cooling systems, which required both trains of Containment Spray to be Operable while in Mode 4. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Condition Prohibited byTechnical Specification and 10 CFR 50.73(a)(2)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function.
The root cause determined that Surveillance Procedure SP 2099, Unit 2 Main Steam Isolation Valve Logic Test, was not adequately designed to account for outage schedule variation. Contributing causes included that the Unit 2 Startup to Mode 4 procedure does not contain adequate process barriers such that plant configuration meets Technical Specification requirements for Mode 4 entry. Operations personnel failed to uphold standards for panel walkdown requirements.
Corrective actions include revising SP 2099, Unit 2 Main Steam Isolation Valve Logic Test to include steps to reposition Containment Spray Switches to the "as found" configuration and revise Unit 2 start-up procedure to add additional HOLD to have the Shift Manager perform Control Board Walkdown to verify equipment required in Mode 4 is aligned and Operable.
Develop and implement an operations improvement plan specifically targeted to improve Operator standards in the performance of Control Board Walkdowns.
NRC FORM 366 (04-2017)
DESCRIPTION OF EVENT
SEQUENTIAL NUMBER
- - 003 REV NO.
- - 00 On November 12, 2017, at 2119, a Control Room board walkdown discovered that both of the Unit 2 Containment Spray Pump control switches [CS-46560 (21 Containment Spray Pump) and CS-46561 (22 Containment Spray Pump)] had been left in pull-out, when operators transitioned Unit 2 from Mode 5 to Mode 4. With the control switches in pull-out, the pumps
~ould not automatically start as required. Technical Specifications (Tech Specs) Limiting Conditions for Operation (LCO) 3.0.3 was entered at 2119 as a result of not complying with LCO 3.6.5, Containment Spray1 and Cooling Systems, which required both trains of Containment Spray to be Operable while in Mode 4. Unit 2 had entered Mode 4 at 0303 on November 12, 2017. LCO 3.0.3 was exited at 2127 on November 12, 2017, when both Containment Spray Pump control switches were placed in automatic restoring Operability. Preliminary investigation determined that while Unit 2 was in Mode 5, Surveillance SP 2099, Main Steam Isolation Valve Logic Test, had taken the Containment Spray Pump control switches to pull-out but did not re-align the control switches to automatic after the test was completed. This was an 8-hour Non-Emergency report per 10 CFR 50. 72(b )(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function.
EVENT ANALYSIS
The event is being reported under 10 CFR 50.73(a)(2)(i)(B), Condition Prohibited by Technical Specification, and 10 CFR 50.73(a)(2)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. This condition meets the reporting criteria because Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was not in full compliance with LCO 3.6.5, Containment Spra/ and Cooling Systems, which required both trains of Containment Spray to be Operable while in Mode 4. Both trains of Containment Spray Pump control switches had been left in pull-out, which defeated the automatic start function for approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and 24 minutes. Entry into Mode 4 while not meeting LCO 3.6.5 was contrary to LCO 3.0.4.
The primary purpose of the Containment Vessel Internal Spray System is to spray cool water into the containment atmosphere, in the event of a loss-of-coolant accident (LOCA) and thereby ensure that containment pressure does not exceed its design value of 46 psig at 268 degrees Fahrenheit.
SAFETY SIGNIFICANCE
A Westinghouse LOCA analysis concludes that the containment design pressure of 46 psig would not have been challenged by the blowdown of a design basis double-ended rupture of the primary loop piping while Prairie Island Unit 2 operated in Mode 4 with both CS pumps locked out. Further, the nominal spray setpoint of 23 psig would not have been reached. Also, two (2) trains of Containment Fan Coolers, operating with service water as high as 95 degrees Fahrenheit (well above actual conditions), would have been able to remove more than enough heat to offset the decay heat rate at 29 days after shutdown.
The Westinghouse Main Steam Line Break (MSLB) analysis also concluded that in the case of a full double-ended rupture MSLB at the SG discharge nozzle, the 46 psig Containment design pressure and the nominal spray setpoint 23 psig valves jwould not have been reached, while Unit 2 operated in Mode 4 with both CS pumps locked out.
The Westinghouse analysis that was performed demonstrated this event did not constitute a Safety System Functional Failure (SSFF). (Reference NEI 99-02, Revision 7, Regulatory Assessment Performance Indicator Guideline, Section 2.2, 1 EllS Component Code-BE Page 2 of 3 (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1 022, R.3 for instruction and guidance for completing this form http :1/vmw. nrc. g ov/read i nq-nn/doc-collections/n u reqs/staff/sr1 022/r3D
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER Prairie Island Nuclear Generating Plant 05000-306 YEAR Unit 2 2017 SEQUENTIAL NUMBER
- - 003 REV NO.
- - 00 Mitigating Systems Cornerstone, Safety System Functional Failures, Purpose.) As such, this event will not be reported in the NRC Performance Indicator (PI) for safety system functional failure since an engineering analysis was performed
~hich determined that the long term Containment pressure would have remained below the design pressure, and automatic actuation of the containment spray system would not have been needed to mitigate the consequences of an accident, even though Prairie Island Unit 2 was not in full compliance with Tech Spec 3.6.5 for the Containment Spray and Cooling Systems.
There were no radiological, environmental, or industrial impacts associated with this event, and PINGP did not adversely affect the health and safety of the public.
CAUSE(S)
Root Cause Surveillance Procedure SP 2099, Unit 2 Main Steam Isolation Valve Logic Test, is not adequately designed to account for outage schedule variation.
Contributing Causes
Startup procedure 2C1.2-M4, Unit 2 Startup to Mode 4, does not contain adequate process barriers such that plant configuration meets LCO requirements for Mode 4 entry.
Operations personnel failed to uphold standards such that the panel walkdown requirements are met to ensure that abnormal conditions and deviations from expected values/positions are communicated evaluated and understood by the Control Room Team.
CORRECTIVE ACTION(S)
SP 2099 will be updated to include steps to reposition Containment Spray Switches to the "as found" configuration.
Revise procedure 2C1.2-M4 to add additional HOLD in section 5.1. 7 to have the Shift Manager perform Control Board Walkdown to verify equipment required in Mode 4 is aligned and Operable.
Develop and Implement an operations improvement plan specifically targeted to improve Operator standards in the performance of Control Board Walkdowns.
PREVIOUS SIMILAR EVENTS
A review of the Corrective Action Program (CAP) and Licensee Event Reports (LERs) for PINGP revealed no results that involved having both control switches in pull out causing a reportable condition. The time frame for this review was eight years.
NRC FORM 3668 (04-2017)