05000327/LER-2013-003, Regarding Limiting Conditions for Operation Exceeded for Emergency Core Cooling System

Regarding Limiting Conditions for Operation Exceeded for Emergency Core Cooling System

ML13301A007
Person / Time
Site:	Sequoyah
Issue date:	10/21/2013
From:	John Carlin Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 13-003-00
Download: ML13301A007 (12)

LER-2013-003, Regarding Limiting Conditions for Operation Exceeded for Emergency Core Cooling System

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
3272013003R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000 October 21, 2013 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 1 Facility Operating License No. DPR-77 NRC Docket No. 50-327 Subject: Licensee Event Report 50-327/2013-003, "Limiting Conditions for Operation Exceeded for Emergency Core Cooling System" The enclosed Licensee Event Report provides details concerning the loss of function for emergency core cooling valves. This report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B), as an event or condition that is prohibited by technical specifications.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Michael McBrearty, Sequoyah Site Licensing Manager, at (423) 843-7170.

Sequoyah Nuclear Plant Enclosure: Licensee Event Report 50-327/2013-003 cc: NRC Regional Administrator-Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant

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

, the NRC may not conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.

3. PAGE Sequoyah Nuclear Plant Unit 1 05000327 1 OF 11

4. TITLE:

Limiting Conditions for Operation Exceeded for Emergency Core Cooling System

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED SEUNILRVFACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR NUMBER NO.

FACILITY NAME DOCKET NUMBER 08 23 2013 2013 -

003 00 10 21 2013

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 LI 20.2201(b)

El 20.2203(a)(3)(i)

LI 50.73(a)(2)(i)(C)

LI 50.73(a)(2)(vii)

[1 20.2201(d)

Ej 20.2203(a)(3)(ii)

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

LI 50.73(a)(2)(viii)(A)

LI 20.2203(a)(1)

[]

20.2203(a)(4)

LI 50.73(a)(2)(ii)(B)

LI 50.73(a)(2)(viii)(B)

_ 20.2203(a)(2)(i)

[: 50.36(c)(1)(i)(A)

E] 50.73(a)(2)(iii)

[1 50.73(a)(2)(ix)(A)

10. POWER LEVEL LI 20.2203(a)(2)(ii)

[L 50.36(c)(1)(ii)(A)

LI 50.73(a)(2)(iv)(A)

LI 50.73(a)(2)(x) 100%

El 20.2203(a)(2)(iii)

[_ 50.36(c)(2)

LI 50.73(a)(2)(v)(A)

LI 73.71(a)(4)

[]

20.2203(a)(2)(iv)

[L 50.46(a)(3)(ii) 50.73(a)(2)(v)(B)

LI 73.71(a)(5)

LI 20.2203(a)(2)(v)

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

LI 50.73(a)(2)(v)(C)

LI OTHER LI 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B)

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

Specify in Abstract below or in B.

Status of structures, components, or systems that were inoperable at the start of the event and contributed to the event:

The 1 B-B Emergency Diesel Generator (DG) [EIIS Code EK] was inoperable for scheduled maintenance. The 11B-B DG provides emergency power to the B-Train of ECCS for design basis events.

C.

Dates and approximate times of occurrences

Dates and Times

Description

August 4, 2013 at 0743 1 B-B Emergency DG made inoperable for scheduled maintenance outage. Unit 1 enters TS LCO 3.8.1.1 Action b.

August 8 at 0709 1-FCV-63-72, A-Train RHR Containment Sump Isolation Valve found with mid-position indication on 1-M-6 hand switch, Panel 6K and the associated local breaker cubicle.

August 8 at 0709 TS LCO 3.3.3.7 is entered for PAM instrumentation of RHR Containment Sump Isolation Valve with an action to restore to operable status within 30 days.

August 8 at 0900 RHR Containment Sump Isolation Valve is declared operable and the action of TS LCO 3.3.3.7 remained in effect.

August 9 at 0040 1 B-B Emergency DG is restored and declared operable. Unit 1 exits TS LCO 3.8.1.1 Action b.

August 9 at 1141 to 1201 1B-B Safety Injection (SI) Pump was made inoperable for casing vent. Unit 1 enters and exits TS LCO 3.5.2 action a.

August 12 at 0042 to 1A-A SI Pump was made inoperable for case venting.

0100 Unit 1 enters and exits TS LCO 3.5.2 action a.

August 12 at 1127 to Commenced depressurizing SI Pump Discharge 1131 piping. Unit 1 enters and exits TS LCO 3.5.2 action a.

ugust 12 at 0552 to 1 B-B RHR Pump is made inoperable for schedule 0555 activities. Unit 1 enters and exits TS LCO 3.5.2 action a.

August 13 at 0000 to 1A-A SI Pump is made inoperable for scheduled work 1210 activities. Unit 1 to enters and exits TS LCO 3.5.2 action a.

August 14 at 0538 A-Train of ECCS flow path is inoperable for maintenance on the centrifugal charging pump injection tank (CCPIT) Inlet Valve 1-FCV-063-0039.

Unit 1 enters TS LCO 3.5.2 action a.

August 14 at 1436 Maintenance on the CCPIT Inlet Valve, 1-FCV-063-0039, is complete. Unit 1 exits TS LCO 3.5.2 action a.

August 14 at 2140 to 1A-A SI Pump is declared inoperable for valve 2154 testing. The 1A-A SI Pump is placed in pull-to-lock.

Unit 1 to enters and exits TS LCO 3.5.2 action a.

Unit 1 enters and exits TS LCO 3.5.2 Action a.

multiple times for scheduled testing of multiple valves.

August 14 at 2315 1A-A RHR Pump is declared inoperable for valve testing. The 1A-A RHR Pump is placed in pull-to-lock. Unit 1 enters TS LCO 3.5.2 Action a.

1A RHR Pump suction valve 1-FCV-74-3 fails to close on demand. Unit 1 remains in LCO 3.5.2 action a and LCO 3.6.2.1.

August 15 at 0904 Clearance number 1-63-1245 (2013) is placed to prevent any inadvertent operation of RHR Containment Sump Isolation Valve while troubleshooting RHR Pump Suction Valve.

August 15 at 1800 Multiple grounds were found in the control circuits for RHR Containment Sump Isolation and RHR Pump Suction Valves.

August 16 at 0315 Water is discovered in the motor limit-switch housing control circuit of RHR Containment Sump Isolation Valve.

August 17 at 0559 Following completion of maintenance activities the A-Train of ECCS is declared operable.

August 23 TVA determined that RHR Containment Sump

Isolation Valve was likely inoperable.

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

1. Motor Operated Valve FCV-063-0072-A RHR Containment Sump Isolation Valve Limitorque Corporation a division of Flowserve Corporation, Actuator, Model Number SMB-3

2. Motor Operated Valve FCV-074-0003-A RHR Pump Suction Valve Square-D Corporation, Relay, Class 8501 Type X E.

Other systems or secondary functions affected

The following components were considered to be adversely effected by the water intrusion of the RHR Containment Sump Isolation Valve limit-switch housing:

1-FCV-072-0040-A - RHR Spray Header A-A Isolation Valve 1-FCV-074-0001-A - RHR System Isolation Valve 1-FCV-063-0008-A - RHR Heat Exchanger Number 1 The normal function of 1-FCV-072-0040-A supports containment isolation. This valve remained closed during this event. Other functions of 1 -FCV-072-0040-A including opening when required for RHR Spray [EIIS Code BE] during a loss-of-coolant accident (LOCA), closing or capable of reclosing to support cold leg recirculation, and isolating a passive failure during hot leg recirculation.

The normal function of the 1 -FCV-074-0001 -A is isolation of the reactor coolant system (RCS) [EIIS Code AB] from the suction side of the RHR pumps. This valve remained closed during this event. The 1-FCV-072-0001-A is required to open for support of cooling when the RCS pressure is below 380 pounds per square inch gauge and shutdown cooling is required.

The normal function of 1-FCV-063-0008-A is to remain closed isolating the suction of the safety injection and centrifugal charging pumps from the discharge of the RHR pump. This valve remained closed during this event. During a LOCA this valve is required to open during containment sump recirculation.

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

1. During the performance of surveillance testing on the RHR Pump Suction Valve, the valve failed to stroke close from its normally open position.

2. Maintenance troubleshooting identified the RHR Containment Sump Isolation Valve to contain multiple grounds, and ultimately determined the limit-switch housing contained water.

G.

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

The closing relay for the RHR Pump Suction Valve had failed. It is postulated that a submerged limit-switch in the RHR Containment Sump Isolation Valve resulted in a high impedance ground allowing a limited voltage application to the closing relay.

The resultant voltage in this circuit path was some value below the relay pickup voltage, allowing the relay coil to remain energized at a reduced value, and insufficient to effect closure of the relay contact. The relay coil being continuously energized may have introduced degradation of the insulation resulting in a short within the coil causing the failure.

Water intrusion into the RHR Containment Sump Isolation Valve limit-switch housing was discovered during maintenance troubleshooting to restore the A-Train of ECCS. The water has been identified as groundwater that permeated through the concrete structure and entered into the electrical conduit that penetrates the ceiling of the RHR Containment Sump Isolation Valve room via unsealed threaded conduit fittings. The valve was likely inoperable due to the water intrusion.

H. Operator actions

Following the discovery on August 8, 2013, of the dual position indication on the RHR Containment Sump Isolation Valve, Operations personnel assessed the condition for operability and entered into the action of LCO 3.3.3.7.

Operations personnel declared the A-Train of ECCS inoperable on August 14, 2013, and entered into the action of LCO 3.5.2, upon the failure of RHR Pump Suction Valve during surveillance testing.

1.

Automatically and manually initiated safety system responses

During the event, plant conditions did not require automatic or manual initiated safety system response.

Ill.

Cause of the event

A.

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

The conduit penetrations for RHR Containment Sump Isolation Valve were not designed to account for groundwater infiltration through the plant concrete structures resulting in leakage into the conduit and subsequently the limit switch compartment.

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

cause

The root cause of this event was determined not to be human performance related.

IV.

Analysis of the event

The design basis of the ECCS is to provide two independent trains of ECCS for accident mitigation. ECCS is designed to cool the reactor core as well as to provide additional shutdown capability following initiation of accident conditions including a pipe break or spurious valve lifting in the RCS which cause a discharge larger than that which can be made up by the normal makeup system, up to and including the instantaneous circumferential rupture of the largest pipe in the RCS; rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident (i.e., RCCA Ejection); a pipe break or spurious valve lifting in the secondary system, up to and including the instantaneous circumferential rupture of the largest pipe in the secondary system; and a steam generator tube rupture (SGTR). The RCS boundary is assumed to remain intact for secondary system events, excluding a SGTR. As such, the failure of the A-Train ECCS RHR Pump Suction and Sump Recirculation Isolation valves would not have impacted these events; therefore, will not be discussed. For the SGTR, the B-Train of the RHR system may not have been available for continued cooldown as a consequence of the 1 B-B DG being inoperable; however, A-Train of the RHR would have been available for long-term cooldown because neither RHR Pump Suction or the Containment Sump Isolation Valves are required to change from their normal position. The remaining events are discussed below.

For a large-break LOCA, the RCS will depressurize rapidly. The safety analysis assumes one complete train of ECCS (centrifugal charging pump [CCP], safety injection pump [SIP], and residual heat removal [RHR] pump and associate flow paths) delivers flow to the reactor coolant system during the injection phase. The recirculation phase is provided for long-term core cooling during the accident recovery period. Both A and B-Train injection flow paths were not always available

during the period of time that valves, RHR Containment Sump Isolation Valve and RHR Pump Suction Valve were failed in closed and open position, respectively, due to ongoing testing and maintenance activities. For a period of time, the 1 B-B Emergency DG was not operable to supply emergency power for the B-Train of ECCS, although normal offsite power was available. Furthermore, for brief amounts of time, the B-Train of ECCS was made inoperable by placing the 1 B-B RHR and SI pumps in pulled-to-lock for scheduled activities. Operators were available to restore the pumps from the Main Control Room. Additionally, the A-Train of ECCS was known to be inoperable by placing various components in an out-of-service condition for maintenance or testing. With each of the valves in a failed position, manual operator action would have been necessary to effect the recirculation phase in the A-Train of ECCS. As a result of the conditions, core cooling may not have been available.

For a small-break LOCA, a much slower depressurization of the RCS will occur.

The safety analysis assumes one complete train of ECCS is available for injection.

As described above, at various times both A and B-Train of ECCS may not have been available during the period of time that the RHR Containment Sump Isolation Valve and the RHR Pump Suction Valve were failed in closed and open position, respectively. Manual operator actions, described in existing emergency operating procedures, provide assurance that core cooling could be mitigated for break sizes up to 2 inches in diameter. For break sizes greater than 2 inches in diameter, operator action may not have been successful in establishing timely recirculation to preclude core damage.

The RCCA ejection accident is initiated by an assumed mechanical failure of the control rod mechanism pressure housing which results in a rapid reactivity insertion.

The responses are rapid with the power spike occurring over a small fraction of a second. The transient, as analyzed, is mitigated by the power range neutron flux low range trip (i.e., for hot zero power) and the power range neutron flux high range trip (i.e., for full power.) A break size up to 2 square inches is assumed as the limiting break. Because this break size is greater than 2 inches in diameter, operator action may not have been successful in establishing timely recirculation to preclude core damage.

V.

Assessment of Safety Consequences

A.

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

Safety-related systems that were needed to remove residual heat may not have been available in the unlikely occurrence of an event. As such, and in

consideration of the analyzed events in the updated final safety analysis, this event resulted in unanalyzed conditions.

A probabilistic risk assessment was performed for the plant configuration surrounding the event time from August 8, 2013, until ECCS was restored operable on August 17, 2013. The result for core damage frequency and large early release frequency are 1.401e-07 and 2.251e-09, respectively.

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:

Once identified on August 14, 2013, the Unit 1 A-Train of ECCS was inoperable, approximately 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> elapsed until the train was returned to operable status.

VI.

Corrective Actions

Corrective Actions are being managed by TVA's Corrective Action Program under problem evaluation report number 772193.

A.

Immediate Corrective Actions

Troubleshooting was performed on the RHR Containment Sump Isolation Valve and RHR Pump Suction Valve to determine the cause of the failures. Necessary components were replaced in each of the valves. A drain was installed in the limit-switch housing of the RHR Containment Sump Isolation Valve and each the conduits leading to the valve were sealed. An inspection of 85 of 111 motor operated valves on Units 1 and 2 for water intrusion was conducted. The inspection excluded valves with installed drains, with conduit leading to the actuator from a lower elevation, and were inaccessible during plant operation. Of these valves, none exhibited signs of water inside the actuator housing.

B. Corrective Actions to prevent recurrence or to reduce probability of similar events occurring in the future:

Redesign the conduit penetrations for each Unit's RHR Containment Sump Isolation Valves to ensure the penetrations are sealed and/or water tight to prevent water from entering the motor operator valve operator associated with each valve.

Maintenance to implement the redesign of the conduit for the Unit 1 A-and B-Train RHR Pump Recirculation Isolation Valves is schedule for the Unit 1 Fall 2013 refueling outage. Because water intrusion has not been identified in the other two RHR Containment Sump Isolation Valves, implementation of the conduit redesign for Unit 2 A-and B-Train RHR Containment Sump Isolation Valves is planned for future refueling outages.

Identify any conduits that penetrate concrete structures below the water table, approximately elevation 690 feet, in locations known to experience water intrusion.

An inspection for water intrusion of a statistically significant population of those penetrations will be conducted, with expansion of the population if water is identified in a conduit. Identified water intrusion into a conduit is to result in the sealing of the conduit.

Perform walk downs (or conduit drawing review) for plant elevation 690 feet and below in locations where water intrusion has been identified to occur. Any conduits or electrical components regarding safety related or trip sensitive equipment, mounted in a location in which they are vulnerable to water intrusion are to be identified. Identified items are to be verified to be properly sealed or that water intrusion would not result in an adverse condition. Identified item not properly sealed or where water intrusion would result in adverse conditions will be evaluated on a case-by-case basis.

VII.

Additional Information

A.

Previous similar events at the same plant:

A review of previous reportable events for the past 3 years did not identify any

B. Additional Information

None.

I

C.

Safety System Functional Failure Consideration:

This event did result in a safety system functional failure.

D.

Scrams with Complications Consideration:

This event did not result in an unplanned scram with complications.

VIII.

Commitments

None