05000296/LER-2002-001, Re Failure to Meet Technical Specification Requirements for Inoperable Primary Containment Isolation Instrumentation Due to Removal of Incorrect Temperature Switches

Re Failure to Meet Technical Specification Requirements for Inoperable Primary Containment Isolation Instrumentation Due to Removal of Incorrect Temperature Switches

ML020850713
Person / Time
Site:	Browns Ferry
Issue date:	03/08/2002
From:	Bhatnagar A Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER-02-001-00
Download: ML020850713 (8)

LER-2002-001, Re Failure to Meet Technical Specification Requirements for Inoperable Primary Containment Isolation Instrumentation Due to Removal of Incorrect Temperature Switches

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)
2962002001R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 Ashok S. Bhatnagar Vice President, Browns Ferry Nuclear Plant March 8, 2002 U.S. Nuclear Regulatory Commission 10 CFR 50.73 ATTN:

Document Control Desk Mail Stop OWFN, P1-35 Washington, D. C.

20555-0001

Dear Sir:

TENNESSEE VALLEY AUTHORITY -

BROWNS FERRY NUCLEAR PLANT (BFN) -

UNIT 3 -

DOCKET 50-296 -

FACILITY OPERATING LICENSE DPR LICENSEE EVENT REPORT (LER) 50-296/2002-001-00 The enclosed report provides details of a failure to meet the requirements of Technical Specification (TS) Limiting Condition for Operation due to inoperability of the primary containment isolation instrumentation for the Reactor Core Isolation Cooling system.

TVA is reporting this event pursuant to 10 CFR 50.73(a)(2)(i)(B), as any operation or condition prohibited by the plant's Technical Specifications. There are no commitments contained in this letter.

Sincerely, Ashok S. Bhatnagar cc:

See page 2 c7 Prited or recycded paper

U.S. Nuclear Regulatory Commission Page 2 March 8, 2002 Enclosure cc (Enclosure):

(Via NRC Electronic Distribution)

Mr. Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. Paul E. Fredrickson, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Resident Inspector Browns Ferry Nuclear Plant P. 0. Box 149 Athens, Alabama 35611

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2001 (6-19981

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

FACILITY NAME 111 DOCKET NUMBER (21 I

PAGE (3)

Browns Ferry Nuclear Plant, Unit 3 05000296 1 of 6 TITLE (41 Failure to Meet Technical Specification Requirements for Inoperable Primary Containment Isolation Instrumentation Due to Removal of Incorrect Temnerature Switches EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME I DOCKET NUMBER NUMBER NUMBER I

NA 05000 DOCKET NUMBER 01 09 2002 2002

-- 001 --

00 03 08 2002 NA l

05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) 111 MODE 9 l

1 20.2201 b)l 20.2203(a)(2)(v) lX 50.73(a)(2)(i)(B) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10 l 100 l 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER

--ll l20.22031a11211iii1 l

50.36(cI~)(1 50.73(a)(2)(v)

Specify in Abstract below or in

}-

1 l20.22031a11211iv1 l

50,36(cI)(2) 50.731a1121(vii) l LICENSEE CONTACT FOR THIS LER (12) lNAME ITELEPHONE NUMBER llnclUde Area Codel IE. Dorris Charlton, Nuclear Engineer, Industry Affairs 1256.729.7533 li COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) llCAUSE lSYSTEM lCOMPONENT lMANUFACTURER lREPORTABLE

- l

CAUSE

ISYSTEM ICOMPONENT IMANUFACTU I

REPORTABLE 11111I TO NPRDS

- i

- lIIl RER I

T PD

.SUPPLEMENTAL REPORT EXPECTED (141 l

EXPECTED iMONTH lDAY f YEAR l lYES lX lNO Ii SUBMISSION l

ll ll l(If yes, complete EXPECTED SUBMISSION DATE),

ll DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) 116)

On January 9, 2002, at 0849 hours0.00983 days <br />0.236 hours <br />0.0014 weeks <br />3.230445e-4 months <br />, four Reactor Core Isolation Cooling (RCIC) steam line space temperature switches were replaced with temperature switches intended for replacement of the High Pressure Coolant Injection (HPCI) steam line space temperature switches. This resulted in the Primary Containment Isolation Instrumentation being inoperable. At 1425 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.422125e-4 months <br />, the Control Room was notified of the error and technical specification actions were initiated to isolate the affected penetration flow path. Since the Primary Containment Isolation instrumentation was made inoperable without the knowledge of Operations, TVA did not take the required technical specification actions within one hour of the instrumentation becoming inoperable. In accordance with 10 CFR 50.73 (a) (2) (i) (B), this report is being submitted as any operation or condition prohibited by the plant's technical specifications The root cause for this event was personnel error. Maintenance personnel failed to properly execute the verification process required by site procedures. The individuals focused only on the address portion of the component identifier rather than on the complete component identifier. Actions to prevent recurrence included appropriate personnel corrective actions for involved personnel and stand-down briefings and refresher training for Instrument and Controls (I&C) maintenance personnel.U.S. NUCLEAR REGULATORY COMMISSION (6-1998(

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME IlI DOCKET LER NUMBER 16 l

PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER 2 of 6 Browns Ferry Nuclear Plant, Unit 3 05000296 2002 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (171 I.

PLANT CONDITION(S)

At the time of the event, Unit 3 was in Mode 1 at 100 percent reactor power, approximately 3458 megawatts thermal. Unit 2 was in Mode 1 at 100 percent reactor power, approximately 3458 megawatts thermal. Unit 1 was shutdown and defueled.

II.

DESCRIPTION OF EVENT

A. Event:

On January 9, 2002, at 0421 hours0.00487 days <br />0.117 hours <br />6.960979e-4 weeks <br />1.601905e-4 months <br /> CST, maintenance personnel (other utility, non-licensed) commenced surveillance 3-SR-3.3.6.1.3(3D), High Pressure Coolant Injection (HPCI) [BJ] Steam Line Space High Temperature Calibration. This surveillance involved replacement of primary containment isolation instrumentation [JM1 temperature switches [TS] with shop-calibrated switches, then returning the removed switches to the shop to obtain as-found calibration data.

At 0849 hours0.00983 days <br />0.236 hours <br />0.0014 weeks <br />3.230445e-4 months <br />, the maintenance personnel removed the primary containment isolation temperature switches for the Reactor Core Isolation Cooling (RCIC) system [BN] and replaced them with the temperature switches intended for the HPCI system. The switches for the RCIC system are the same type switches and located in close proximity to the HPCI switches. However, the actuation temperature for the HPCI switches is 170 degrees F increasing instead of the 147 degrees F increasing required for the RCIC system. Similar component identifiers are used which differ only by the system number (e.g. 3-TS-071-0002N (RCIC) and 3-TS-073-0002N (HPCI)).

While preparing to obtain the as-found calibration data, it was discovered that the RCIC steam line space temperature switches had been replaced instead of the HPCI switches as intended.

At 1425 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.422125e-4 months <br />, maintenance personnel notified control room operators that the incorrect temperature switches had been replaced. Operators initiated technical specification action for inoperable primary containment isolation instrumentation; within one hour, isolate the affected penetration flow path(s).

At 1509 hours0.0175 days <br />0.419 hours <br />0.0025 weeks <br />5.741745e-4 months <br />, operators completed the technical specification required action to isolate the RCIC steam supply line by closing valve 3-FCV-71-3 [FCV]. RCIC was declared inoperable and Technical Specification LCO 3.5.3.A.2 was entered.

At 1818 hours0.021 days <br />0.505 hours <br />0.00301 weeks <br />6.91749e-4 months <br />, after replacement of the RCIC temperature switches, Operations personnel exited LCO 3.3.6.1.F.1.

At 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br />, after returning RCIC to service, Operations personnel exited Technical Specification LCO 3.5.3.A.2.

Because the Primary Containment Isolation Instrumentation was inoperable for a period longer than that allowed by Technical Specifications, this event is reportable in accordance with 10 CFR 50.73(a) (2) (i)

(B), as any operation or condition prohibited by plant's Technical Specifications.

B. Inoperable Structures. Components, or Systems that Contributed to the Event:

None NRC FORM 366 16-1998)U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1 )

DOCKET LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER 3 of 6 Browns Ferry Nuclear Plant, Unit 3 05000296 2002 --

001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

C. Dates and Approximate Times of Masor Occurrences:

January 9, 2002, at 0421 hours0.00487 days <br />0.117 hours <br />6.960979e-4 weeks <br />1.601905e-4 months <br /> CST January 9, 2002, at 0849 hours0.00983 days <br />0.236 hours <br />0.0014 weeks <br />3.230445e-4 months <br /> CST January 9, 2002, at 1425 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.422125e-4 months <br /> CST January 9, 2002, at 1509 hours0.0175 days <br />0.419 hours <br />0.0025 weeks <br />5.741745e-4 months <br /> CST January 9, 2002, at 1818 hours0.021 days <br />0.505 hours <br />0.00301 weeks <br />6.91749e-4 months <br /> CST January 9, 2002, at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br /> CST Maintenance personnel began performance of surveillance 3-SR-3.3.6.1.3(3D), HPCI Steam Line Space High Temperature Calibration.

Maintenance personnel incorrectly replaced temperature switches for the RCIC steam line space high temperature instead of for the HPCI system as required by the surveillance.

Maintenance personnel notified Operations personnel of incorrect temperature switch replacement.

Operations personnel initiated technical specification LCO action 3.3.6.1.F.1 to manually isolate the RCIC steam line.

Operations personnel completed the technical specification LCO action. RCIC system was declared inoperable and Technical Specification LCO 3.5.3.A.2 was entered.

Operations personnel exited Technical Specification LCO action 3.3.6.1.F.1 after satisfactory replacement of the RCIC temperature switches.

Operations personnel declared RCIC system operable and exited Technical Specification LCO 3.5.3.A.2.

D. Other Systems or Secondary Functions Affected

None

E. Method of Discovery

While preparing to perform the as-found portion of the calibration surveillance, it was determined that the incorrect switches were replaced. Control room operators were notified of the error at 1425 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.422125e-4 months <br />.

F. Operator Actions

Upon notification of the error, Operations personnel initiated technical specification LCO action 3.3.6.1.F.1, manually isolate the RCIC steam line. The RCIC system was declared inoperable and Technical Specification LCO 3.5.3.A.2 was entered.

G. Safety Svstem Resoonses None NRC FORM 366 (6-1998)U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE (3)

YEAR S EQUENTIAL REVISION NUMBER 4 of 6 Browns Ferry Nuclear Plant, Unit 3 05000296 2002 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (171 Ill.

CAUSE OF THE EVENT

A. Immediate Cause Maintenance personnel removed the temperature switches for the RCIC system instead of the HPCI switches as required by 3-SR-3.3.6.1.3(3D).

B. Root Cause The root cause of this event was personnel error. The Maintenance personnel involved failed to properly execute the verification process in accordance with approved site procedures. The personnel focused on the "address" portion of the unique identification number (UNID) rather than the complete UNID.

C. Contributing Factors HPCI and RCIC switches are of the same type and located in close proximity to each other. The UNIDs for the HPCI and RCIC temperature switches are identical with the exception of the system identification number. (e.g. 3-TS-071-0002N (RCIC) and 3-TS-073-0002N (HPCI))

IV.

ANALYSIS OF THE EVENT

Surveillance Requirement 3-SR-3.3.6.1.3(3D), High Pressure Coolant Injection Steam Line Space High Temperature Calibration, requires the HPCI steam line space temperature switches 3-TS-073-0002N, -0002P,

- 0002R, and -0002S to be replaced with calibrated switches. Procedure steps for removal and replacement of the temperature switches require first and second party verification. Procedure SPP-10.3, "Verification Program" requires the second-party verifier to ensure that the actual component identification matches the identification of the component required to be verified. The maintenance personnel did not properly implement error prevention techniques such as self-checking. The verifications focused on only the address portion of the UNID instead of the entire UNID. This resulted in the incorrect switches being replaced.

This was the first time that these particular individuals had performed this task. The planned access route to the area was not accessible and another route was used. This resulted in the individuals becoming disoriented with respect to the correct work location. The similarity and proximity of the HPCI and RCIC switches and the similarities in their UNIDs led the individuals to believe that they were at the correct location.

When commencing the portion of the surveillance to obtain the as-found data for the removed switches, it was discovered that the incorrect switches had been replaced. Operations was notified of the error and initiated the required Technical Specification actions.

V.

ASSESSMENT OF SAFETY CONSEQUENCES

The function of RCIC steam space temperature switches 3-TS-071-0002N, -0002P, -0002R, and -0002S is to monitor for postulated breaks in the piping supplying motive steam to the RCIC system turbine. The Browns Ferry Updated Final Safety Analysis Report Section 7.3.4.7 states:

High temperature in the vicinity of the RCIC equipment could indicate a break in the RCIC steam line. The automatic closure of certain Group A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier. When high temperature occurs near the RCIC equipment, the RCIC NRC FORM 366 (6-1998)U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER 16)

PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER 5 of 6 Browns Ferry Nuclear Plant, Unit 3 05000296 2002 --

001 00 TEXT (If more space is required, use additional copies of NRC Form 366A} (17) turbine steam line is isolated. The high temperature isolation setting was selected far enough above anticipated normal RCIC system operational levels to avoid spurious operation, but low enough to provide timely detection of an RCIC turbine steam line break.

In this case the "Group A" valves referenced are the RCIC steam line isolation valves, which are signaled to close upon detection of a possible leak from the RCIC steam line piping.

Through the installation of additional supports, the RCIC steam line piping in the torus area has been made particularly rugged. The piping stresses seen therefore by the RCIC steam line piping in the areas monitored by these temperature switches are very low.

Upon confirmation that the incorrect switches had been replaced, the RCIC steam line was isolated in accordance with the technical specification required LCO action. This was completed within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of the original error. The correct switches were reinstalled and demonstrated to be operable within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of the original error.

The RCIC temperature switches which were removed had a nominal actuation setpoint of 147 degrees F increasing, whereas the HPCI temperature switches which were installed in their place had a nominal actuation setpoint of 170 degrees F increasing. The installed switches were functional in this location, though the isolation of the RCIC steam line upon a postulated RCIC steam line pipe break in the area would have occurred at a higher temperature. This could have resulted in a greater loss of steam from the reactor pressure boundary into secondary containment prior to isolation.

The following factors are seen to reduce the risk impacts of this event:

1. The BFN design basis has demonstrated that the RCIC steam piping is exposed to very low stresses in the areas monitored by these temperature switches, making the likelihood of a pipe break very small

2. The incorrect switches were in place for only a short period of time prior to taking the appropriate LCO action (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />)

3. During this short interval, the isolation function would have occurred for a postulated pipe break, though at a higher isolation temperature

4. The BFN risk-informed in-service inspection (RI-ISI) program estimated the total risk contribution of all RCIC system piping (water and steam) breaks to be only 0.02% of the BFN total CDF.

Based on the above discussion, the safety effects of this event are considered to be negligible.

V.

CORRECTIVE ACTIONS

A. Immediate Corrective Actions Initiated and implemented work order to replace, functionally test, and return to service the RCIC temperature switches.

B. Corrective Actions to Prevent Recurrence Administered appropriate personnel corrective actions to involved personnel.

Conducted stand-down briefings on each shift with Instrument and Controls (I&C) Maintenance personnel to stress the proper verification methods and importance of following those methods.

Refresher verification training was administered to I&C Maintenance personnel.

NRC FORM 366 16-1998)U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE 13)

YEAR SEQUENTIAL REVISION NUMBER 6 of 6 Browns Ferry Nuclear Plant, Unit 3 05000296 2002 --

001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

VI.

ADDITIONAL INFORMATION

A. Failed Components None B. Previous LERs on Similar Events None

C. Additional Information

None D. Safety System Functional Failure:

This event did not result in a safety system functional failure in accordance with NEI 99-02.

VIL.

COMMITMENTS

None NRC FORM 366 (6-1998)