05000410/LER-2015-003

LER-2015-003, Primary Containment Isolation Function for some valves not maintained during Surveillance Testing

Nine Mile Point Unit 2
Event date:	6-23-2015
Report date:	3-31-2016
Reporting criterion:	10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material
4102015003R01 - NRC Website

LER 15-003-01 for Nine Mile Point, Unit 2, Regarding Primary Containment Isolation Function For Some Valves Not Maintained During Surveillance Testing

ML16102A278
Person / Time
Site:	Nine Mile Point
Issue date:	03/31/2016
From:	Kreider R E Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NMP2L 2618 LER 15-003-01
Download: ML16102A278 (6)

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A. PRE-EVENT PLANT CONDITIONS:

Prior to the event, Nine Mile Point Unit 2 (NMP2) was operating at 100 percent power.

B. EVENT:

This event was discovered on June 23, 2015 during a management review.

On April 22, 2015, the Primary Containment Isolation function for the Reactor Building Closed Loop Cooling and Reactor Water Cleanup valves was inadvertently defeated for 47 minutes during surveillance testing.

On May 8, 2015, the valves for the opposite division in the same systems were similarly impacted for 50 minutes during surveillance testing of the opposite division.

This resulted in an inadvertent loss of the Reactor Vessel Water Level (Level 2) Primary Containment Isolation Safety Function.

Nine Mile Point Unit 1 (NMP1) was unaffected by the surveillance at NMP2.

The event has been entered into the station's corrective action program as IR 2518177.

C. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

No other systems or secondary functions were affected beyond the systems discussed in Section I.B. There were no safety system actuations during this time frame.

II. CAUSE OF EVENT:

The Surveillance testing on April 22 was for valves powered by Division 2. To prevent an inadvertent full isolation signal from occurring during the testing on Division 2, the power supply breakers for the Division 2 valves was opened while they were being tested. The plan was to prevent a system actuation during testing, but ensure the ability to maintain isolation capability because the Division 1 side was not impacted.

During work preparation activities, an error was made resulting in the work order steps for opening the power supply breakers to prevent inadvertent closure of the primary containment isolation valves.

Personnel did not recognize that performing the surveillance on two of the trip units for Division 2 would result in losing the closure capability for the opposite division (Division 1) outboard primary containment isolation valves. This resulted in a loss of automatic primary containment isolation ‘113C FORM 366A U.S. NUCLEAR REGULATORY COMMISSION '02-2014)

CONTINUATION SHEET

r.

• 1 function on Reactor Vessel Low Water level, Level 2, during the time of the testing of the two trip units.

The same error was made on May 5 during Division 1 testing.

III. ANALYSIS OF THE EVENT:

This event is reportable under 10 CFR 50.73 (a)(2)(v)(C) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

On the two dates above, both containment isolation valves (inboard and outboard) for those systems would not have closed if an automatic Reactor Vessel Low Water Level 2 Isolation signal was generated. One division of valves had the breakers open preventing closure, while the other division Level 2 Isolation was bypassed for testing.

There were no actual nuclear safety consequences associated with this event. The consequences of the event were mitigated by the diversity of trip initiation for pipe breaks inside the primary containment being provided by monitoring drywell high pressure and associated circuitry for providing a primary containment isolation signal to isolate the valves on a Drywell High Pressure signal.

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

This event constitutes a safety system functional failure. The NRC Reactor Oversight Process indicator for Safety System Functional Failures at Unit 2 will increase from 0 to 1 and remain green.

IV. CORRECTIVE ACTIONS:

A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL

STATUS:

The plant was returned to pre-event normal status after the completion of each surveillance.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

The procedures for performing the surveillance have been revised to ensure the correct power supply breakers are opened for each section of the surveillance.

A. FAILED COMPONENTS:

There were no failed components that contributed to this event.

B. PREVIOUS LERs ON SIMILAR EVENTS:

There are no similar LERs for NMP2.

C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION

COMPONENT

IEEE 803 FUNCTION IEEE 805 SYSTEM

IDENTIFIER IDENTIFICATION

Primary Containment Isolation Valve

D. SPECIAL COMMENTS:

None