05000333/LER-2011-001, Regarding Reactor Core Isolation Cooling System Inoperable Longer than Allowed by Technical Specifications

Regarding Reactor Core Isolation Cooling System Inoperable Longer than Allowed by Technical Specifications

ML110670304
Person / Time
Site:	FitzPatrick
Issue date:	03/08/2011
From:	Bronson K Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-11-0029 LER 11-001-00
Download: ML110670304 (6)

LER-2011-001, Regarding Reactor Core Isolation Cooling System Inoperable Longer than Allowed by Technical Specifications

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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)(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
3332011001R00 - NRC Website

text

~Entergy JAFP-11-0029 March 8, 2011 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 TeI315-342-3840 Fax 315-349-6480 Kevin Bronson Site Vice President - JAF

SUBJECT:

Dear Sir or Madam:

LER: 2011-001-00, Reactor Core Isolation Cooling System Inoperable Longer Than Allowed By Technical Specifications James A. FitzPatrick Nuclear Power Plan Docket No.

50-333 License No.

DPR-59 This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications..."

There are no commitments contained in this report.

Questions concerning this report may be addressed to Mr. Joseph Pechacek, Licensing Manager, at (315) 349-6766.

sin~e7' p

/'~~~

Kevin Bronson Site Vice President KB/JP/ed

Enclosure:

JAF LEA: 2011-001-00, Reactor Core Isolation Cooling System Inoperable Longer Than Allowed By Technical Specifications cc:

USNRC, Region 1 USNRC, Project Directorate USNRC, Resident Inspector INPO Document Components:

001 Transmittal Letter with Enclosure

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)

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 ted back to industry.

Send comments regarding burden estimate to the FOIAIPrivacy Section (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to LICENSEE EVENT REPORT (LER) infocoliectsJesource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If ameans used to impose an informaflon collection does not display acurrently 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 James A. FitzPatrick Nuclear Power Plant 05000333 1 OF 5

4. TITLE Reactor Core Isolation Cooling System Inoperable Longer Than Allowed By Technical Specifications

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED YEARISEQUENTIALIREV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO MONTH DAY YEAR 09 23 2010 2011 03 08 2011 FACILITY NAME DOCKET NUMBER 001 00

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) r-r-

r-f-

20.2201 (b) 20.2203(a)(3)(i) f-50.73(a)(2)(i)(C)

I-50.73(a)(2)(vii)

Mode 01 F-20.2201 (d) 20.2203(a)(3)(ii) l-50.73(a)(2)(ii)(A) l-50.73(a)(2)(viii)(A)

I-20.2203(a)(1 )

20.2203(a)(4) l-50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B)

I-20.2203(a)(2)(i) 50.36(c)(1 )(i)(A)

I-50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)

10. POWER LEVEL I-20.2203(a)(2)(ii)

- 50.36(c)(1 )(ii)(A) l-50.73(a)(2)(iv)(A) 50.73(a)(2)(x) f-20.2203(a)(2)(iii)

I-50.36(c)(2) 50.73(a)(2)(v)(A) 73.71 (a)(4) l-I-

20.2203(a)(2)(iv)

F-50.46(a)(3)(ii)

I-50.73(a)(2)(v)(B) 73.71 (a)(5) 100 F-20.2203(a)(2)(v)

'= 50.73(a)(2)(i)(A)

I-50.73(a)(2)(v)(C)

OTHER L-20.2203(a)(2)(vi) 129 50.73(a)(2)(i)(B)

~

50.73(a)(2)(v)(D)

Specify in Abstract below or in

EVENT DESCRIPTION

2011 -

001 00 On January 7,2011, with the plant operating in Mode 1 at 100% power, 13MOV-131, Reactor Core Isolation Cooling (RCIC) System [BN] Steam Admission Isolation Valve failed to stroke full open during surveillance test ST-24J, RCIC Flow Rate and Inservice Test.

Troubleshooting determined the most probable cause to be loose connections in the 125 VDC [EJ] motor control circuit. Loose connections were identified going to contactors 42-10, 42-20, and 42-2C, in motor control circuit 71 BMCC-3-0B1 (MC). Based on the strip chart recording, taken during the performance of ST-24J, the observed failure was the result of de-energizing the 42-20 contactor coil. When the 42-20 contactor coil de-energized, the seal-in contact maintaining power to the 42-10 and 42-20 contactors and a main contact in the DC motor series field opened. This fully de-energized the motor operator prior to fully opening the steam admission valve (13MOV-131).

Preventive maintenance (PM) had been performed on the motor control circuit on September 23, 2010, during Refueling Outage (RO) 19. A similar failure occurred on October 29, 2010, during ST-24J. At that time the failure was attributed to a lack of stem lubrication causing the motor torque switch to open on high torque. Motor operated valve 13MOV-131 stroked properly following stem cleaning and lubricating and the system satisfactorily passed the post maintenance test and subsequent surveillance testing. The actual cause of the deficiency was masked by 1) the intermittent nature of the opening of the circuit; and 2) the proper response after the valve stem was lubricated.

When the problem occurred again on January 7,2011, a new Failure Modes and Effects Analysis (FMEA) and further trouble shooting of the control circuit was performed. The FMEA and troubleshooting identified the loose connections to the motor control contactors.

The loose connections were un-lugged compression type connections. This particular style of terminal connection uses a base with a center tapped inverted u-shaped retention plate tightened by a screw inserted through the center tap. Leads are inserted under the retention plate on either side of the screw and the connection is tightened by torquing the screw such that the leads are secured between the base and the retention plate. At the time of discovery the leads associated with the 42-10, 42-20, and 42-2C contactor coils were determined to be loose, and the lead to the 42-20 contactor coil could be removed from the terminal with minimal force.

The PM performed on September 23, 2011, requires checking leads for tightness. This is accomplished by hand-checking the tightness by gently manipulating the wire. The PM does not require a tightness check of the termination using a screw driver or other tool.

The identified condition could have resulted in a failure of the RCIC system to operate properly, if needed.

Because RCIC is not required to be Operable during Mode 2 (Start-up) until steam dome pressure is greater than 150 psig, it is considered that the RCIC System was Inoperable from the time that RCIC was required to be Operable on October 16, 2010, until the completion of the Post Maintenance Testing on January 8, 2011.

Since LCO 3.5.3 requires RCIC to be Operable in Mode 1 and in Modes 2 and 3 with steam dome pressure greater that 150 psig, this period of inoperability exceeded the Technical Specification allowed out of service time_

NflC f OF1M 366A (10-2010)(10~201D)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE James A. FitzPatrick Nuclear Power Plant 05000333 YEAR I SEQUENTIAL I REV I--__--'--_N_U_M_B_ER_--'-_N_O_.---I 3 OF 5

BACKGROUND:

2011 -

001 00 The RCIC System is comprised of various components which include pumps, valves, piping, and instrumentation. The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level under these conditions. The High Pressure Coolant Injection (HPCI) [BJ] and RCIC systems perform similar functions. The HPCI and RCIC systems permit the plant to be shutdown while maintaining sufficient reactor vessel water inventory until the reactor vessel pressure is low enough to allow the Low Pressure Coolant Injection (LPCI) System [BO] or Core Spray (CS) System [BM] to maintain core cooling.

13MOV-131 is the RCIC Steam Admission Isolation valve. The motor operator for 13MOV-131 is supplied power from the 125 VDC Electrical Distribution System.

Although the RCIC System is considered to have been Inoperable, during the period between, October 16, 2010, when the system was first required to be Operable, and the satisfactory performance of post maintenance testing on January 8, 2011, the system was available. The determination of availability is based on the satisfactory performance of surveillance testing (ST-24J) after lUbricating the valve stem, and the ability to manually open 13MOV-131, if system operation had been required.

EVENT ANALYSIS

During James A. FitzPatrick's 2010 refueling outage, a PM was performed on 13MOV-131. At the conclusion of the PM activity it appears that the connections to motor control contactors 42-10,42-20, and 42-2C were loose.

The loose connections resulted in an intermittent open circuit to the 42-20 contactor which would cause the motor operator to stop prior to fUlly opening 13MOV-131. Due to the intermittent nature of the problem it is possible that RCIC might have failed to operate as designed under accident conditions.

Since the TS LCO 3.5.3 requires the system to be Operable in Mode 1 and in Modes 2 and Mode 3 with steam dome pressure greater that 150 psig, it is was concluded that the system was Inoperable for a period of time greater than allowed by the Technical Specifications. However, since the JAF accident analysis does not credit the operation of the RCIC system to mitigate the consequences of a design basis accident there was no loss of safety function.

CAUSE OF EVENT

The most probable cause of this event was loose connections in motor control circuit 71 BMCC-3-0B1 (MC), for motor operator 13MOV-131 (OP).

EXTENT OF CONDITION:

An Extent of Condition (EOC) review for possible loose connections inadvertently missed during Motor Control Center (MCC) PMs during RO-19, similar to 71 BMCC-3-0B1 (MC) for 13MOV-131 was considered.

The EOC review considered 1) Motor Control Centers in which PM activities were performed; 2) which PMs110-2010!

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE James A. FitzPatrick Nuclear Power Plant 05000333 YEAR I SEQUENTIAL I REV t--__.........._N_U_M_B_ER_---L._N_O_.--I 4 OF 5 EXTENT OF CONDITION (continued):

2011 -

001 00 included the same check of connections; and 3) valve position indications obtained during surveillance testing subsequent to the PMs.

Five (5) similar PM activities were identified. The PM activities were performed on the MCC's associated with the following Motor Operated Valves (MOVs): 23MOV-19 and 23MOV-16 in the HPCI System and 13MOV-39, 13MOV-14 and 13MOV-132 in the RCIC System.

STAN, HPCI Quick Start, Inservice, and Performance Monitoring Test, was performed on October 13,2010 and on January 18, 2011. During the performance of these tests 23MOV-16 and 23MOV-19 were required to operate and no problems with valve positioning or indication were identified.

ST-24J, RCIC Flow Rate and Inservice Test, was performed on October 29,2010, and on January 7,2011. In both cases 13MOV-39, 13MOV-14 and 13MOV-132 operated normally with no positioning or indication problems identified. However, as noted in this LER, 13MOV-131 was identified as having positioning and indication problems during both tests. After tightening the loose connections and retesting on January 8, 2011, all the RCIC valves in the EOC population functioned normally.

Because multiple surveillance tests were performed involving the EOC population and no similar indications of loose connections were observed, the extent of the deficiency was determined to be confined to 13MOV-131.

FAILED COMPONENT IDENTIFICATION:

Manufacturer:

Manufacturer Model Number:

NPRDS Manufacturer Code:

NPRDS Component Code:

FitzPatrick Component ID:

CORRECTIVE ACTIONS

Completed Actions:

General Electric IC28001607F3 G080 42 71 BMCC-3-0B1 (MC)

1. Tightened the connections in 71 BMCC-3-0B1 (MG).

2. Tested the System.

Open Actions:

1. Review the apparent cause evaluation with the Electrical Maintenance Department, focusing on inspection of termination practices.

2. Benchmark industry practice on un-lugged wires in compression type fitting terminations to determine if methodology / procedure changes are required.

3. Include this event and any beneficial practices indentified during benchmarking in continuing training for Electrical and Instrument and Control Maintenance.

NRC "ORM 366A (10-2010)(10-2010!

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

2. DOCKET YEAR I SEQUENTIAL I REV I--__-'-_N_U_M_B_ER_----L.._N_O_"---t 5 OF 5

1. FACILITY NAME James A. FitzPatrick Nuclear Power Plant

ASSESSMENT OF SAFETY CONSEQUENCES

Actual Consequences 05000333 2011 -

6. LER NUMBER 001 00

3. PAGE There were no actual industrial, radiological, or nuclear safety consequences during or as a result of the described period of RCIC inoperability.

Potential Consequences The identified condition could have resulted in a failure of the RCIC system to operate as designed. However, the JAF accident analysis does not take credit for the operation of the RCIC system. The accident analysis assumes the operation of the HPCI system and operation of the Automatic Depressurization System (ADS) in conjunction with the Low Pressure Coolant Injection System, which were unaffected by the identified condition.Also, while the automatic operation of the RCIC system was affected by the identified deficiency, the system was available for use by manually opening 13MOV-131 locally. Therefore, the potential consequences of the RCIC system being inoperable during this period were minimal.

SIMILAR EVENTS

Licensee Event Reports (LERs) written since 2000 were reviewed to determine if there were similar events at JAF. LER-2006-002 documented a case of the HPCI System being inoperable for longer than allowed by the plant Technical Specifications due to hydraulic tubing being incorrectly re-installed after a PM activity. Although the 2006 event was not related to an electrical connection it was related to a PM activity.

No other events since 2000 had any similarities with the event reported in this LEA.

REFERENCES:

1.

Apparent Cause Evaluation Report CR-JAF-2011-00123 (RCIC Steam Admission Valve Failed to Open) 2.

Technical Specification 3.5.3 3.

JAF Updated Final Safety Analysis Report Section 4.7, Reactor Coolant Isolation Cooling System