05000259/LER-2018-004-01, Hiqh Pressure Coolant Injection Declared Inoperable Due to Steam Supply Valve Isolation

Hiqh Pressure Coolant Injection Declared Inoperable Due to Steam Supply Valve Isolation

ML19015A071
Person / Time
Site:	Browns Ferry
Issue date:	01/15/2019
From:	Hughes D Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 2018-004-01
Download: ML19015A071 (11)

LER-2018-004, Hiqh Pressure Coolant Injection Declared Inoperable Due to Steam Supply Valve Isolation

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(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)
2592018004R01 - NRC Website

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Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 January 15, 2019 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 1 Renewed Facility Operating License No. DPR-33 NRC Docket No. 50-259

Subject:

Licensee Event Report 50-259/2018-004-01 10 CFR 50.73

Reference:

Letter from TVA to NRC, "Licensee Event Report 50-259/2018-004-00,"

dated September 7, 2018 Pursuant to 10 CFR 50.73(a)(2)(v)(A), (8), and (D), Browns Ferry Nuclear Plant hereby submits supplemental Licensee Event Report (LER) 50-259/2018-004-01 which documents the inoperability of the Browns Ferry Nuclear Plant, Unit 1, High Pressure Coolant Injection system. The enclosed LER has been revised to document a revision of the Level 2 Apparent Cause Evaluation which was initiated at the time when the referenced report was submitted to the NRC.

There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact M. W. Oliver, Acting Nuclear Site Licensing Manager, at (256) 729-7874.

Enclosure: Licensee Event Report 50-259/2018-004 High Pressure Coolant Injection Declared Inoperable due to Steam Supply Valve Isolation

U.S. Nuclear Regulatory Commission Page 2 January 15, 2019 cc (w/ Enclosure):

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

ENCLOSURE Browns Ferry Nuclear Plant Unit 1 Licensee Event Report 50-259/2018-004-01 High Pressure Coolant Injection Declared Inoperable due to Steam Supply Valve Isolation See Enclosed

NRC FORM 366 (04-2018)

1. Facility Name U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020

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

2. Docket Number

3. Page Browns Ferry Nuclear Plant, Unit 1 05000259 1 OF 8

4. Title Hiqh Pressure Coolant Injection Declared Inoperable due to Steam Supply Valve Isolation

5. Event Date

6. LER Number

7. Report Date

8. Other Facilities Involved Month Day Year Year I Sequential I Rev Number No.

Month Day Year Facilrty Name N/A Docket Number N/A 07 09 2018 2018 -

004 01 01 15 2019 Facility Name N/A Docket Number N/A

9. Operating Mode 11. This Report is Submitted Pursuant to the Requirements of 10 CFR §: (Check all that apply)

D 20.2201(b)

D 20.2203(a)(3)(i)

D 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A) 1 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)

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)(2)(ii)

D 50.36(c)(1)(ii)(A)

C8J 50.73(a)(2)(v)(A)

D 73.71(a)(4)

D 20.2203(a)(2)(iii)

D 50.36(c)(2)

C8J 50.73(a)(2)(v)(B)

D 73.71(a)(5)

D 20.2203(a)(2)(iv)

D 50.46(a)(3)(ii)

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

D 73.77(a)(1) 100 D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

C8J 50.73(a)(2)(v)(D)

D 73.77(a)(2)(i)

D 20.2203(a)(2)(vi)

D 50.73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 73.77(a)(2)(ii)

D 50.73(a)(2)(i)(C)

D OTHER Specify in Abstract below or in B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event There were no structures, systems, or components (SSCs) whose inoperability contributed to this event.

C. Dates and approximate times of occurrences

Dates & Approximate Times July 9, 2018, 0920 CDT July 9, 2018, 0958 CDT July 9, 2018, 1111 CDT July 9, 2018, 1253 CDT July 9, 2018, 1310 CDT July 10, 2018, 1342 CDT Occurrence MIG commenced 1-SR-3.3.6.1.2(3B) - HPCI System Steam Supply Low Pressure Functional Test.

Isolation valves closed.

Isolation valves were found closed. HPCI declared inoperable and determined to have been inoperable since 0958 CDT. Entered TS LCO 3.5.1 Condition C.

HPCI declared available, but inoperable. Exited 1-AOl-64-2B.

MIG satisfactorily completed 1-SR-3.3.6.1.2(3B) -

HPCI System Steam Supply Low Pressure Functional.

HPCI declared operable. Unit 1 exited TS LCO 3.5.1 Condition C.

D. Manufacturer and model number of each component that failed during the event

Two PSs were identified as the potential points of failure (1-PS-073-0001A and 1-PS-073-0001C). Both were manufactured by Static-0-Ring (SOR). The manufacturer part number is 5N6-B3-U8-C1A-JJTTNQ. The electrical portion of the SOR PS used in the HPCI steam supply low pressure logic is a Honeywell Micro Switch brand microswitch, part number 11SM244.

E. Other systems or secondary functions affected

No other systems or secondary functions were affected by this event.

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

The unexpected Unit 1 HPCI system isolation was discovered by the Unit Operator on observation of the Control Room panel. Light indications in the Control Room include the position indication lights for the HPCI inboard steam isolation valve and the HPCI outboard steam isolation valve. Additionally, the lamps for Group 4 of the Containment Isolation Status System would have illuminated when the isolation signal was present.

REV NO.

01

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

The failure mechanism was determined to be electrical contact degradation due to loading near the manufacturer's rated values for the microswitch in the HPCI steam supply low PS.

This resulted in momentary arcing during functional testing of an adjacent switch, which energized the logic relays and caused Unit 1 HPCI to isolate.

H. Operator actions

Upon receipt of the BFN, Unit 1, PCIS Group 4 Isolation:

Entered Abnormal Operating Instruction 1-AOl-064-0002B, Group 4 High Pressure Coolant Injection Isolation.

Declared BFN, Unit 1, HPCI inoperable and unavailable and entered TS LCO 3.5.1, ECCS - Operating Condition C.

Verified RCIC was operable.

I.

Automatically and manually initiated safety system responses

During this event, momentary arcing of a PS resulted in a PCIS Group 4 isolation of the HPCI system. PCIS Group 4 isolation closes the steam supply valves to HPCI in the event of high steamline space temperature, high steam flow, or low steamline pressure. These signals are indicative of a line break in the HPCI system steamline to the turbine or high pressure between the diaphragm rupture discs on the HPCI Turbine Exhaust. At the time of the event, these conditions did not exist.

Ill.

Cause of the event

A. Cause of each component or system failure or personnel error The most likely cause of the unexpected Unit 1 HPCI isolation during performance of the HPCI steam supply low pressure functional test was operation of the switch near the manufacturer's electrical contact current ratings. This was a result of a design change that added logic relays to the isolation circuit, increasing the current flow through the circuit.

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

No human performance related root causes were identified.

REV NO.

01 IV.

Analysis of the event

The HPCI system is provided to assure that the reactor is adequately cooled to limit fuel cladding temperature in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the nuclear plant to be shut down, while maintaining sufficient reactor vessel water inventory until the reactor vessel is depressurized. The HPCI system continues to operate until the reactor vessel pressure is below the pressure at which Low Pressure Coolant Injection (LPCl)[BO] operation or Core Spray System operation maintains core cooling. Due to the isolation of the HPCI system, this system would have been unable to perform its safety function.

The design discrepancy that resulted in the unexpected isolation is the addition of logic relays to the HPCI steam supply low pressure isolation circuit. The design configuration that has been evaluated in Tab C-3 of EQ Binder BFNOEQ-IPS-003 as acceptable for the PS electrical contacts includes two GE HFA type relays in parallel being energized by the PS contact closure. Since the time that the initial evaluation was performed, an Agastat EGP type relay was added to the Unit 1 circuit, and two GE HGA type relays were added to the Units 2 and 3 circuits. The additional current demanded on the circuit to energize the relays has resulted in approaching the switch electrical contact rating. Electrical contacts degrade in repeated "make" and "break" cycles by pitting, which can affect contact performance and reliability. The pitting is a result of material transfer across contacts that occurs from arcing on either contact make or break. In DC circuits, the pitting can be more severe than in AC circuits, as one contact is always positive and the other is always negative. This pitting is exacerbated in inductive loading applications by the counter electromotive force which causes high break currents.

The electrical portion of the SOR switch used in the HPCI steam supply low pressure logic is a Honeywell Micro Switch brand microswitch, part number 11 SM244. The switch is robust enough to continue to operate acceptably with some contact degradation; however, the two instances of unexpected HPCI isolations during the PS functional testing demonstrates that the degradation has affected the switches in a manner that can occasionally result in momentary spurious closure.

A voltmeter used in the functional testing procedure detects that the switch contacts are open, but the voltmeter test voltage across the contacts is very low. However, when approximately 280VDC is applied across the contacts, a momentary arc across the air gap of the contacts, which can be affected by the pitting, can result in momentary circuit completion and energization of the logic relays. The failure mechanism has been difficult to detect because of the intermittent and unpredictable frequency.

The potential for spurious system isolation with the HPCI steam supply low PSs in standby is low based on the logic configuration (one-out-of-two-taken-twice). The configuration is changed during quarterly functional testing as one-half of the one-out-of-two-taken-twice logic is satisfied while each switch is being tested. This can be changed by a minor modification to the testing procedure to lift a wire for the switch under test to isolate it from the logic circuit. This is acceptable because TS Table 3.3.6.1-1, Primary Containment Isolation Instrumentation, provides 01 in Function 3.b for the HPCI Steam Supply Line Pressure - Low, that only 3 channels out of 4 are required to remain operable.

V.

Assessment of Safety Consequences

This event resulted in inoperability and unavailability of the single train of the BFN, Unit 1, HPCI system resulting in the inability of the HPCI system to perform its safety functions for shutting down the reactor and maintaining it in a safe shutdown condition, mitigation of the consequences of an accident, and the removal of residual heat in the event that the reactor was shut down. In the event of an emergency, the RCIC system remained operable, and all ADS systems were available during this event to facilitate core cooling by low pressure ECCS systems. Based on the above, during the time period that the HPCI system was inoperable, sufficient systems were available to provide the required safety functions to protect the health and safety of the public.

There was no significant reduction to the health and safety of the public or plant personnel for this event.

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, RCIC was verified as operable by Operations personnel. Additionally, all other ECCS and ADS systems remained operable.

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 shutdown.

C. For failure that rendered a train of a safety system inoperable, estimate of the elapsed time from discovery of the failure until the train was returned to service The event resulted in isolation of the BFN, Unit 1, HPCI system at 0958 CDT on July 9, 2018.

The event was discovered at 1111 CDT on July 9, 2018, at which point Unit 1 was retroactively declared inoperable by Operations personnel. However, HPCI remained inoperable until 1342 CDT on July 10, 2018. The lapsed time for inoperability of the HPCI system was 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> and 44 minutes.

VI.

Corrective Actions

Corrective Actions are being managed by TV/\\s corrective action program under Condition Reports (CRs) 1429497 (1-FCV-73-2 found closed during board walkdown) and 1437895 (Discrepancy in HPCI PS contact rating).

01

A. Immediate Corrective Actions

The immediate corrective actions related to this event were to replace the two Unit 1 pressure switches that could have failed, resulting in spurious isolation. Based on the one-out-of-two-taken-twice logic configuration, it was not determined which Unit 1 PS actually failed, but it was narrowed down to two of the four switches. Work orders were also initiated to replace the other two Unit 1 PSs, as well as Unit 2 PSs and Unit 3 PSs.

B. Corrective Actions to Prevent Recurrence or to reduce the probability of similar events occurring in the future The PS functional test procedures for all three units, 1/2/3-SR-3.3.6.1.2(3B), have been revised to reduce the chances of inadvertent isolation by lifting a wire for the PS under test and then re-landing the wire prior to moving on to the next PS. Lifting a wire for the PS under test eliminates the purposeful making up of one-half of the one-out-of-two-taken-twice logic, such that a single failure will not result in spurious actuation. In addition to lifting the wires, CR 1437895 was initiated to document the discrepancy identified with the electrical contact rating for the SOR brand PSs used in the HPCI steam supply low pressure isolation circuit.

CR Action 1429497-05 is being used to monitor the corrective actions of CR 1437895 to ensure resolution of the nonconforming condition related to the HPCI steam supply low pressure isolation switches.

VII.

Previous Similar Events at the Same Site

A search of BFN Licensee Event Reports (LERs) for Units 1, 2, and 3, identified five LERs similar to this issue within the last six years.

BFN LER 260/2012-004-00 -- High Pressure Coolant Injection Rendered Inoperable Due to an Inadvertent Actuation of Primary Containment Isolation System (PER 596706)

BFN LER 259/2013-007-00 -- High Pressure Coolant Injection Declared Inoperable Due to an Inadvertent Actuation of the Primary Containment Isolation System (PER 794807)

BFN LER 296/2015-004-00 -- High Pressure Coolant Injection Inoperable Due to Failed Pressure Switch (CR 1024825)

BFN LER 296/2016-006-00 -- High Pressure Coolant Injection Found Inoperable During Testing (CR 1179483)

BFN LER 259/2016-002-00 -- High Pressure Coolant Injection Safety System Functional Failure due to lnoperability of Primary Containment Isolation Valve (CR 1193943)

A review of the corrective actions for the CRs associated with the LE Rs concluded that the corrective actions associated with these CRs would not have prevented this event from occurring.

01

VIII. Additional Information

There is no additional information.

IX.

Commitments

There are no new commitments. -- t>T UMts: NU. 31ou-ui04 CA,..lr<c;:,: U~l~ll.tU.tU Estima1ed, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

YEAR 2018

3. LER NUMBER SEQUENTIAL NUMBER 004 REV NO.

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