05000245/LER-1997-018-01, :on 970224,four MOVs May Not Fully Isolate Downstream Pipe Break.Caused by Lack of Adequate Design Change Process Guidance & Limited Vendor & Industry Documentation Availability.Mov Program Will Be Revised

:on 970224,four MOVs May Not Fully Isolate Downstream Pipe Break.Caused by Lack of Adequate Design Change Process Guidance & Limited Vendor & Industry Documentation Availability.Mov Program Will Be Revised

ML20137H509
Person / Time
Site:	Millstone
Issue date:	03/26/1997
From:	Robert Walpole NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20137H479	List: 05000245/LER-1997-018, Forwards LER 97-018-00,documenting an Event That Occurred at Millstone Nuclear Power Station,Unit 1 on 970224.Util Commitments Made within Ltr,Listed 05000245/LER-1997-018-01, :on 970224,four MOVs May Not Fully Isolate Downstream Pipe Break.Caused by Lack of Adequate Design Change Process Guidance & Limited Vendor & Industry Documentation Availability.Mov Program Will Be Revised
References
LER-97-018-01, LER-97-18-1, NUDOCS 9704020265
Download: ML20137H509 (4)

LER-1997-018, on 970224,four MOVs May Not Fully Isolate Downstream Pipe Break.Caused by Lack of Adequate Design Change Process Guidance & Limited Vendor & Industry Documentation Availability.Mov Program Will Be Revised

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
2451997018R01 - NRC Website

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l (See reverse for required number of digits / characters for each block) 1 FACILITY NAME (1)

DOCKET NUMBER (2)

PAGE (J)

Millstone Nuclear Power Station Unit 1 05000245 1 of 4 TITLE 14)

Four MOVs May Not Fully isolate Pipe Breaks EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8) 1 i

MONTH-DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACluTY NAME E UCKET NUMBER NUMBER I

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02 24 97 97 018 00 03 26 97 l

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OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

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20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) l 50.73(a)(2)(viiil POWER 20.2203(aH1) 20.2203(a)(3Hi)

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20.2203(aH3Hii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2Hii) 20.2203(aH4) 50.73(a)(2)(iv)

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LICENSEE CONTACT FOR THIS LER (12) l NAME TELEPHONE NUMBER unclude Area Code) j Robert W. Walpole, MP1 Nuclear Licensing Manager (860)440-2191 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

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ABSTRACT (Limrti s i.e., approximately 15 single-spaced tynewntten Iones) (16) 4 On February 24, 6,, with the plant in COLD SHUTDOWN, a review of the as-left limit switch settings of valves that have a torque switch bypass feature indicated that four valves may not fully isolate a downstream pipe break. These four valves have a spring pack compensator which allows the stem to move upward during the closing valve stroke.

j This movement was not accounted for in determining the physical location on the valve seat where the valve disc would stop. The result is that the disc may not be covering the valve seats, thus not isolating the flow through the valve, before the bypass of the torque switch is electrically removed and the valve is assumed to stop. These valves 11-j IC-1,1-lC-4,1-CU 2 and 1-CU-28) isolate the Isolation Condenser system and the Reactor Water Cleanup systems rcapectively.

This condition was immediately reported as outside the design basis of the plant on February 24,1997. The cause of the event is the lack of adequate design change process guidance and limited vendor and industry documentation 3

availability.

To prevent reoccurrence, a design change to compensate for the potential upward movement of the stemnut and 4

stsm during a high energy line break condition is required. This will include adjusting the close torque switch bypass parcentage setting in the field. This corrective action will be completed prior to startup for operating Cycle 16, 9704020265 970326 PDR ADOCK 05000245 i

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• • U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LEO NUMbM (6)

PAGE 13)

T YEAR SEQUENTIAL HEVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 2 of 4 97 018 00 TEXT (11more space is required, use additional ccpies of NRC Form 366A) (17)

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Descrintion of Event On Febmary 24,1997, with the plant in COLD SHUTDOWN, a review of the as-left limit switch settings of valves that have a torque switch bypass feature indicated that four valves may not fully isolate a downstream pipe break.

These four valves have a spring pack compensator which allows the stem to move upward during the closing j

vrive stroke. This movement was not accounted for in determining the physicallocation on the valve seat where ths valve disc would stop. The result is that the disc may not be covering the valve seats, thus not isolating the flow through the valve, before the bypass of the torque switch is electrically removed and the valve is assumed to l

stop. These valves (1-IC-1,1-lC-4,1-CU 2 and 1-CU-28) isolate the Isolation Condenser system and the Reactor Water Cleanup systems respectively.

11.

Cause of Event

The cause of the event is the lack of MOV program manual guidance and limited vendor and industry documentation availability. This resulted in the misappucation and inadequate design of the motor operated vrives (MOV) torque switch bypass circuit and limit switch settings. As a result, this caused the design differences between the two models of MOV to be missed. The effects of the spring compensator unit design on the proposed change was also not fully understood by the plant and other MOV project engineers during the dssign, development, and incorporation of the torque switch bypass circuit.

Ill. Analvsis of Event Th3 four subject vakes are inboard isolation condenser outlet isolation valve 1-IC-1, isolation condenser inboard raturn isolation valvc 1 IC-4, reactor water cleanup inboard inlet isolation valve 1-CU-2 and reactor water cleanup outboard outlet isolation valve 1-CU-28. These valves isolate the isolation condenser and the reactor water i

cizznup systems respectively. These valves are considered Generic Letter 89-10 Supplement 3 valves, which are valves that are designed to isolate piping line breaks. The valves also serve as containment isolation valves for Loss of Coolant Accidents (LOCA) inside the drywell.1-lC-1,1-IC-4 and 1-CU-2 are located inside the Drywell.1-l CU-28 is located outside the Drywell.

Thsse four valves have a spring pack compensator which allows the stem nut to move upward during the closing valve stroke. The valves have Limitorque 'SB' model operators. The movement of the stem nut results in the final stem travel not to be as f ar into the valve body. This movement was not accounted for in determining the physical location on the valve seat where the valve disc would stop.

The control circuits for several MOVs at Millstone Unit No.1 have been designed to bypass the torque switch for>95% of the closing stroke. This means that the torque switch is bypassed for almost all of the valve travel.

This bypass feature allows the motor to exert all available torque output to shut the valve against the large differential pressures created during a pipe break transient. During the closing stroke, once the valve opening port is covered by the valve disk, the torque switch is electrically re-inserted into the circuit at a predetermined position j

in the valve seats.

i When the torque switch is back in the circuit, the valve stops because the torque switch has exceeded its setpoint.

This is of no concern for Limitorque "SMB" actuators that do not incorporate the stem nut spring compensator 1

tssembly, as the valve's opening port would be covered to isolate flow. However, in the case of the Limitorque

• SB" models, the result of the upward movement of the stem nut and thus the stem against the spring compensator unit during the closing stroke may result in the disc not covering the valve seats and not fullyI l -

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• • NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION i

Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 3 of 4 97 018 00 TEXT (If more space is required. use additional copies of NRC Form 366A) (17) isolating flow through the valve. Of the several MOVs designed with the >95% close bypass feature, only 4 of -

thzse MOVs are "SB" Units (which are the 4 described above).

t Tha consequences of the failure of the valves to fully close off flow in the event of a High Energy Line Break (HELB) is that there would still be minor flow through the valves. Analysis of LOCA events outside of containment are bounded by the break in the recirculation sample lines and a break in the reactor water cleanup system. This entlysis assumes that the break is isolated. The safety evaluation for the installation of the design change for the i

torque switch bypass assumed that there would be flow cutoff, but that there may be some leakage as the valve may not be fully scated.

It is assumed that the leakage that would result from this new condition could exceed that which was assumed.

I From static diagnostic test data obtained during past refueling outages, and estimating the closing forces required to isolate flow (based on a conservative valve factor and worst case differential pressure across the disk during the closing stroke), it was determined that the gate valves could be as much as 3/8" from flow cutoff when the i

torque switch is reinserted into the closing circuit.

With the torque switches for all four valves set at a lower torque setting than that predicted under worst case conditions, the valves could stop short and not achieve flow cutoff. This type of leakage would exceed the design basis assumptions relative to the reactor water cleanup i

system and isolation condenser system release rates and volume.

Current e.mergency operating procedures would direct the operator to depressurize the reactor to reduce the

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leakage.

Tha other design feature for these valves is that they are containment isolation valves and maintain containment intsgrity for LOCA events in the drywell. LOCA events in the drywell are not postulated to occur concurrently with HELB events. Therefore, the isolation valves do not see the same type of flow loads during a containment Isolation signal as they would during a HELB event. Under the conditions for containment isolation, the spring pack compensator does not see the upward thrust and therefore the valves are expected to achieve the full travel and full seating. The valves would therefore perform their safety function for containment isolation for events inside the drywell. In addition, the inaoility to ensure primary containment leakage would remain within the maximum Appendix J leakage requirements L., as the leakage would be contained by the drywell.

IV. Corrective Action

To prevent reoccurrence, a design change to compensate for the potential upward movement of the stemnut j

and stem during a high energy line break condition is required.

TV. will include adjuating the close torque switch bypass percentage setting in the field This corrective arsn will be completed prior to startup for operating Cycle 16.

i Additionally, the Continuing Training topics will be updated to include the function and operation of the compensator spring pack in Limitorque *SB" type actuators to all mechanics and technicians associated with work on Limitorque actuators, i

The MOV Program Manual Section PI-5 will be revised to address the torque switch bypass implications for the Limitorque "SB" type actuators.-

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' *U.S. NUCLEAR REGULQTORY COMMISSION (44s) l LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

I YEAR SEQUENTIAL REVISION Millstone' Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 4 of 4 i

97 018 00 TEXT (If more space is required, use additionalcopies of NRC form 366A) (17) i V.

Similar Events

i LER 94-019, "1-MS-5 & 6 and 1-FW-4A and B Unable to perform design basis function pre-FRO-14".

i LER 94-025, " Isolation Condenser Containment isolation Valve Timing Failure",

Manufacturer Data I

1-lC-1 is a 14" crane solid wedge gate valve with a SB-3 Limitorque Actuator 1-IC-4 is a 10" Crane solid wedge gate valve with a SB-2 Limitorque Actuator 1-CU-2 is a 8" Crane solid wedge gate valve with a SB-1 Limitorque Actuator l

1-CU-28 is a 8" Crane solid wedge gate valve with a SB-2 Limitorque Actuator

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