05000296/LER-2014-003, Regarding Primary Containment Isolation Valve Inoperable for Longer than Allowed by Technical Specifications

Regarding Primary Containment Isolation Valve Inoperable for Longer than Allowed by Technical Specifications

ML14217A340
Person / Time
Site:	Browns Ferry
Issue date:	07/31/2014
From:	Polson K Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 14-003-00
Download: ML14217A340 (9)

LER-2014-003, Regarding Primary Containment Isolation Valve Inoperable for 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
2962014003R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 July 31, 2014 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 3 Renewed Facility Operating License No. DPR-68 NRC Docket No. 50-296

Subject:

Licensee Event Report 50-29612014-003-00 The enclosed Licensee Event Report provides details of the inoperability of a Browns Ferry Nuclear Plant, Unit 3, Primary Containment Isolation Valve. The Tennessee Valley Authority (TVA) is submitting this report in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(i)(B), as any operation which was prohibited by the plant's Technical Specifications.

There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact J. L. Paul, Nuclear Site Licensing Manager, at (256) 729-2636.

pectful J. P s

o Site ice Presi~dent

Enclosure:

Licensee Event Report 50-296/2014-003 Primary Containment Isolation Valve Inoperable for Longer than Allowed by Technical Specifications cc (w/ Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant

ENCLOSURE Browns Ferry Nuclear Plant Unit 3 Licensee Event Report 50-296/2014-003-00 Primary Containment Isolation Valve Inoperable for Longer than Allowed by Technical Specifications See Enclosed

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 01131/2017 (01-2014)

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

3. PAGE Browns Ferry Nuclear Plant, Unit 3 05000296 1 of 7

4. TITLE: Primary Containment Isolation Valve Inoperable for Longer than Allowed by Technical Specifications

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEARTSQETA E

MONTH DAY YEAR NANA NNUMBER NO.

N/A N/A FACILITY NAME DOCKET NUMBER 06 02 2014 2014 -

003 -

00 07 31 2014 N/A N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b)

[I 20.2203(a)(3)(i)

El 50.73(a)(2)(i)(C)

[I 50.73(a)(2)(vii)

[1 20.2201(d)

[I 20.2203(a)(3)(ii)

Cl 50.73(a)(2)(ii)(A)

E] 50.73(a)(2)(viii)(A)

[1 20.2203(a)(1)

El 20.2203(a)(4) 0l 50.73(a)(2)(ii)(B)

[] 50.73(a)(2)(viii)(B)

El 20.2203(a)(2)(i)

El 50.36(c)(1)(i)(A)

El 50.73(a)(2)(iii)

[I 50.73(a)(2)(ix)(A)

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

[I 50.36(c)(1)(ii)(A) 0l 50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x)

[: 20.2203(a)(2)(iii)

[I 50.36(c)(2)

El 50.73(a)(2)(v)(A)

El 73.71(a)(4)

El 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

[I 50.73(a)(2)(v)(B)

El 73.71 (a)(5) 100 El 20.2203(a)(2)(v)

[I 50.73(a)(2)(i)(A)

El 50.73(a)(2)(v)(C)

[I OTHER El 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B)

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

Foin A

12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (include Area Code)

Eric Bates, Licensing Engineer 256-614-7180AUMANU-REPORTABLE NMANU-REPORTABLE

CAUSE

SYSTEM COMPONENT FANU-REPOA

CAUSE

SYSTEM COMPONENT FACTURER TO EPIX FACTURER TO EPIX FCUE OEI N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR SUBMISSION E: YES (If yes, complete 15. EXPECTED SUBMISSION DATE)

[

NO DATE N/A N/A ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On June 2, 2014, during performance of the Reactor High Pressure Calibration surveillance, the Residual Heat Removal (RHR) Shutdown Cooling (SDC) Inboard Suction Valve Isolation relay failed to energize preventing automatic closure of the RHR SDC Inboard Suction Valve. On three occasions, the inability of this valve to close automatically upon receipt of the Primary Containment Isolation System signal resulted in a violation of the Browns Ferry Nuclear Plant, Unit 3, Technical Specifications. The Shutdown Cooling Mode of the Residual Heat Removal System was unaffected by this condition.

The cause of the event was relay wires had been lifted and incorrectly landed due to a human performance error at an indeterminate time between a successful post maintenance test (PMT) on March 07, 2014, and the time the condition was corrected by re-landing the wires according to plant drawings on June 6, 2014.

The corrective action to reduce likelihood of recurrence is to develop and deliver a case study to the Maintenance, Modifications, and Operations departments based on the details of this event.

NRC FORM 366 (02-2014)

C. Dates and approximate times of occurrences

March 2, 2014 The RHR SDC Inboard Suction Valve Isolation relay was replaced during the Unit 3 refueling outage16.

March 7, 2014 Satisfactory functioning of the RHR SDC Inboard Suction Valve Isolation relay was confirmed by Post Maintenance Test.

June 2, 2014, at 0430 CDT The RHR SDC Inboard Suction Valve Isolation relay failed to actuate during performance of the Reactor High Pressure Calibration surveillance.

June 6, 2014 Electrical Maintenance corrected the wire terminations per work instructions, returning functionality of the relay.

D. Manufacturer and model number (or other identification) of each component that failed during the event:

No component failures were identified that occurred during the event.

E. Other systems or secondary functions affected

There were no other systems or secondary systems affected.

F. Method of discovery of each component or system failure or procedural error

On June 2, 2014, at 0430 CDT, the Shutdown Cooling Inboard Suction Valve Isolation relay failed to actuate during performance of the Reactor High Pressure Calibration surveillance.

G. The failure mode, mechanism, and effect of each failed component, if known:

There were no failed components related to this event; however, the RHR SDC Inboard Suction Valve Isolation relay failed to actuate due to improperly landed wires.

H. Operator actions

At time of discovery, the Unit 3 Unit Supervisor entered Technical Specification Section 3.3.6.1.A, which Required Action A.1 requires placing the channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I. Automatically and manually initiated safety system responses

There were no automatic or manual safety system responses associated with this event.

1

II1.

Cause of the event

A. The cause of each component or system failure or personnel error, if known:

The most likely cause of this event involves inappropriate manipulation of the RHR SDC Inboard Suction Valve Isolation relay leads by individuals intending to perform scheduled work on another component. Once the leads were lifted the performers realized the error and re-landed the leads incorrectly. As long as the two wires were landed together the connection to ground would be intact and the RHR SYS I Logic Power Failure alarm would not be sealed in. Also no alarm would be received if a jumper was installed. If this event occurred while other relays were being tested in the outage, receiving the alarm would not be unusual and would be attributed to the ongoing work in the Auxiliary Instrument Room. Although no one self-reported such an error, this type of error most closely matches the evidence collected during the investigation and is the most likely cause of the event and thus is the apparent

cause

B. The cause(s) and circumstances for each human performance related root

cause

The relay wires being landed on the wrong terminal is a human performance event.

However, the exact time and nature of the error or the type of individuals involved could not be determined. Between the date of the PMT and the failed surveillance, no approved plant process documentation was found that directed the relay leads to be lifted.

IV.

Analysis of the event

On June 2, 2014, during performance of the Reactor High Pressure Calibration surveillance, the RHR SDC Inboard Suction Valve Isolation relay failed to energize. The relay's failure to energize, preventing automatic closure of the RHR SDC Inboard Suction Valve. The inability to close automatically upon receipt of a Primary Containment Isolation System Group 2 isolation signal is a Technical Specifications (TS) violation.

In cases where both the RHR SDC Inboard Suction Valve and the RHR SDC Outboard Suction Valve were open, the affected penetration flowpath was not isolated. A review of the timeline between relay replacement and discovery identified that the RHR SDC Inboard Suction Valve and the RHR SDC Outboard Suction Valve were both in the open position for greater than one hour on three occasions (to support shutdown cooling operations) as summarized below.

March 10, 2014, at 0500 CDT to March 13, 2014, at 0903 CDT - a total of 76.05 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (Mode 5 / Mode 4)

March 13, 2014, at 2218 CDT to March 17, 2014, at 0034 CDT - a total of 74.27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> (Mode 4)

May 7, 2014, at 0138 CDT to May 7, 2014, at 1846 CDT - a total of 17.13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> (Mode 3)

The requirements for Primary Containment Isolation Instrumentation are in TS section 3.3.6.1. TS Limiting Condition for Operation (LCO) 3.3.6.1 states that the primary

containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.

At time of discovery, the Unit 3 Unit Supervisor entered TS Section 3.3.6.1.A, which Required Action A.1 requires placing the channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Functions 6.b and 6.c and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Function 6.a.

TS Table 3.3.6.1-1 notes that Function 6.a (High Reactor Steam Dome Pressure) and Function 6.c (High Drywell Pressure) are applicable in Modes 1, 2, and 3. On two of the three occasions listed above, the reactor was in Mode 4 or Mode 5. In the third occasion, the reactor was in Mode 3. Therefore, Functions 6.a and 6.c were applicable during the time period from 0138 CDT to 1846 CDT on May 7, 2014. During this time period, Required Action F.1 to isolate the affected penetration flowpath was not met, and the subsequent Required Actions of Condition G are to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. As the reactor was in Mode 3 when the Condition was entered, and the Condition was exited within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, the TS requirements were met for Functions 6.a and 6.c.

For Function 6.b (Low Reactor Vessel Water Level - Level 3), Required Action A. 1 is to place the channel in trip in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. When this was not met, Condition C required entering the referenced condition in Table 3.3.6.1-1. Table 3.3.6.1-1 notes that Function 6.b is applicable in Modes 3, 4, and 5, and the referenced condition is Condition I. The Required Actions for Condition I are (1. 1) to immediately initiate action to restore the channel to operable status or (1.2) to immediately initiate action to isolate the RHR Shutdown Cooling System. As the wiring error was not identified between March 7, 2014, and June 2, 2014, no actions were initiated to restore the channel to operable or maintain the RHR Shutdown Cooling System isolated during this time period.

Therefore, Required Action 1.2 was not met each time RHR Shutdown Cooling was not isolated and the TS were not met for the three occasions listed in above.

Based on the above analysis, the Tennessee Valley Authority (TVA) is submitting this report in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(i)(B), as any operation which was prohibited by the plant's Technical Specifications.

V. Assessment of Safety Consequences

This event resulted in inoperability of the BFN, Unit 3, RHR SDC Inboard Suction Valve and failure to perform its safety function to be able to close automatically upon receipt of a PCIS Group 2 isolation signal. The function of the RHR SDC Inboard Suction Valve Isolation relay is to energize upon receipt of a Group 2 PCIS signal on High Reactor Steam Dome Pressure, High Drywell Pressure, or Low Reactor Water Level. Energizing the RHR SDC Inboard Suction Valve Isolation relay initiates system logic that closes the RHR SDC Inboard Suction Valve and inhibits its opening.

The Shutdown Cooling mode of RHR was not affected by the condition as the automatic closure function is not required for the Shutdown Cooling safety function. The RHR SDC Inboard Suction Valve is required to be open for Shutdown Cooling. The improperly wired relay did not affect the ability of the operator to open or close the valve using the remote hand switch.

Three Primary Containment Isolation Instrumentation functions (High Reactor Steam Dome Pressure, High Drywell Pressure, and Low Reactor Water Level) for the RHR SDC Inboard Suction Valve were affected by the improper wiring of RHR SDC Inboard Suction Valve Isolation relay. However, in accordance with the RHR operating instruction's Precautions and Limitations, when Reactor Vessel Pressure is greater than atmospheric pressure the RHR SDC Cooling Suction Outboard Valve is required to remain closed with its breaker OFF, except for testing or shutdown cooling operation.

This is an Appendix R requirement which minimized the time that the affected penetration flow path (SDC suction line) was not in an isolated condition.

In addition to the above, the RHR SDC Suction Outboard Valve provides redundant primary containment isolation in the affected penetration flow path (SDC suction line).

The primary containment isolation instrumentation for automatic closure of the RHR SDC Outboard Suction Valve receives the same three signals as the instrumentation for the RHR SDC Inboard Suction Valve, but the automatic closure of RHR SDC Outboard Suction Valve uses a different relay. Because the wiring error on RHR SDC Inboard Suction Valve Isolation relay did not affect the SDC Outboard Suction Valve Isolation relay, isolation capability of the RHR SDC Outboard Suction Valve was not affected.

Therefore, the primary containment flowpath would have isolated on an actual isolation signal.

Based on the discussion above, the safety significance of this event is minimal and the event did not pose a threat to the health and safety of the public or plant personnel.

A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event:

During this event, the RHR SDC Inboard Suction Valve remained closed and the RHR SDC Outboard Suction Valve remained operable as the redundant PCIV.

B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident:

This event did not occur when the reactor was shut down.

C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from discovery of the failure until the train was returned to service:

This event resulted in inoperability of the BFN, Unit 3, RHR SDC Inboard Suction Valve to automatically close for approximately 4 days from discovery of the failure until the relay was rewired and returned to service.

VI.

Corrective Actions

Corrective Actions are being managed by TVA's corrective action program under Problem Evaluation Report (PER) 892500.

Immediate Corrective Actions

Performance of the Reactor High Pressure Calibration surveillance was halted.

A Work Order was executed to corrected wiring of the RHR SDC Inboard Suction Valve Isolation relay.

Reactor High Pressure Calibration surveillance was performed satisfactorily.

Corrective Actions Reduce Probability of Similar Events Occurring in the Future The corrective action to reduce the likelihood recurrence is to develop and deliver a case study to the Maintenance, Modifications, and Operations departments based on the details of this event.

VII.

Additional Information

A. Previous similar events at the same plant:

A review of relevant TVA Operating Experience including PERs and LERs was conducted based on the initial details known about the event. Typically the search is expanded once the cause has been identified. In this instance, the specific details of the cause was not identified since the exact time and nature of the error or the type of individuals involved could not be determined. However, the relay wires being landed on the wrong terminal is a human performance event. Numerous examples were found regarding wires being landed on the wrong contacts. Generally, the corrective actions were to re-land the wires per the print and administer discipline to the individual(s) involved.

However, due to the specific cause of this event being indeterminate, no correlation could be made as to if the learnings from those events would have or should have prevented this event.

B. Additional Information

There is no additional information.

C. Safety System Functional Failure Consideration:

In accordance with Nuclear Energy Institute (NEI) NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," this event is not considered a safety system functional failure. The RHR SDC Outboard Suction Valve remained available to perform the isolation safety function.

D. Scram with Complications Consideration:

This event did not result in a reactor scram.

VIII. COMMITMENTS

There are no commitments.