ML19354D826

LER 89-029-00:on 891213,determined That Max Actuator Torque Switch Setting Insufficient to Overcome Friction Force Due to Age Hardening of Elastomeric Seat Matl on Bif/General Signal Corp Valve.Also Reported Per Part 21.W/900112 Ltr

ML19354D826
Person / Time
Site:	Catawba
Issue date:	01/15/1990
From:	Glover R, Owen T DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-PT21-90, REF-PT21-90-010-000 LER-89-029-01, LER-89-29-1, PT21-90-010-000, PT21-90-10, NUDOCS 9001220210
Download: ML19354D826 (8)

Text

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DUKEPOWER January 12, 1990-Document Control Desk U. S.-Nuc1 car Regulatory Commission, Washington, D. C. 20555

Subject:

Catawba Nuclear Station-Docket ~No. 50-413 LER 413/89-29 Gentlemen:

. Attached'is Licensee Event-Report 413/89-29. submitted as a Courtesy.? ,

Report'concerning POTENTIAL INOPERABILITY OF COMPONENT COOLING-ISOLATION VALVES DUE-TO AGE-HARDENED ELASTOMERIC SEAT MATERIAL.z This event was. considered to be of no significance with respect to the-health and safety of'the public.

cry trul yours,

'ITony B. Owen Station Manager keb: COURTESY LER xc: Mr. S. D. Ebneter American Nuclear Insurers _ ,

Regional Administrator, Region II c/o Dottie'Sherman, ANI. Library i U. S. Nuclear Regulator Commission The Exchange,: Suite 245 101 Marietta Street, NW, Suite 2900 270 Farmington Avenue Atlanta, GA 30323 Farmington, CT 06032 M & M Nuclear Consultants Mr. K; Jabbour 1221 Avenues of the Americas U. S. Nuclear Regulatory. Commission' How York, NY 10020 6ffice of Nuclear Reactor-Regulation-Washington, D. C. 20555 INPO Records Center Suite 1500 Mr. W. T. Orders 1100 Circle 75 Parkway NRC Resident Inspector' R Atlanta, GA 303.39 Catawba Nuclear Station j j

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... LICENSEE EVENT REPORT (LER) l . ACiuTv AME m ooCxET NousER m Paar a Catawba Nuclear Station, Unit 1 01510l0l0l41113 1 lOFl Ol 7 n T ' Potential Inoperability of Component Cooling-Isolation Valves 7 Due To Age-Hardened Elastomeric Seat Material EVENT DATE Ili LER NUMeER (s) REPORT OATE 171 OTHER F ACILITIES INVOLVID (8)

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ARIA;OQt R.M. Glover, Compliance Manager 81013 81311 l- l 3121316 COMPLETE ONE LINE 70R E ACM COMPONENT F AILURE DESCRISED IN THIS REPORT t131 CAUSE SYSTEM COMPONENT MA A? A OmTA MA C. m " t g g qpq CAUSE SY ST E M COMPONENT 0 NPR B CI C VI I i B 12 1 5 l 0 Y I i i i l I l I l i I I I I I I I I I I I SUPfLEMENTAL REPORT EXPECTED 114: MONTM DAY YEAR

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1 Operation, Compliance issued a Technical Specification Operability Notification identifying valves IKC-81B, KC to ND Hx 1B Supply Isolation, and 2KC-56A, KC to ND Hx 2A Supply Isolation, as inoperable. At that time, both Units 1 and 2 entered the Technical Specification Action Statement for one inoperable train of the Emergency Core Cooling System. A potential inoperability was' identified by Design Engineering after determining that the maximum actuator torque switch setting may not be sufficient to overcome the friction force that may exist due to age hardening of the elastomeric seat material. These KC System valves have never failed to open when called upon during performance tests, thus it was a conservative approach to determine them " inoperable" and make the necessary torque switch adjustments. The actuator's open torque switch was bypassed for I fifty percent of the valve stroke. The actuator is sufficiently rated to produce the required opening torque with the torque switch bypassed. Units 1 and 2 exited their action statements on December 14, 1989, at 0244 hours0.00282 days <br />0.0678 hours <br />4.034392e-4 weeks <br />9.2842e-5 months <br /> and 1 0325 hours0.00376 days <br />0.0903 hours <br />5.373677e-4 weeks <br />1.236625e-4 months <br />, respectively. This incident is classified as a manufacturer's ,

functional design deficiency due to an inadequate estimate of the degree of age '

hardening in the elastomeric seat material. This report is being submitted as a l Courtesy LER and is reportable pursuant to 10CFR21. I 1

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F BACKGROUND L

i The Residual Heat Removal [EIIS:BP] (ND) System functions as part of the. i y Emergency Core Cooling System (ECCS). The'ECCS1is actuated by a Safety j Injection. Signal with' the flow alignment changing over time after initiation, j The ND System can be used to deliver cooling water directly to the Reactor j

, Coolant [EIIS:AB] (NC) System provided that NC~ pressure has dropped bel.ow:the ND "

pump's discharge pressure. The ND System-can also provide water to the suction -

of other ECCS pumps capable of injecting into the NC. System at higher pressures.

Heat is removed from the ND System via two ND Heat Exchangers [EIIS:HX] (Hx). ,

The Component-Cooling [EIIS:CC) (KC) System essential header provides cooling water to the ND Hxs.

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The KC Supply Isolation Valves [EIIS:V] to the ND Mxs are.normally closed and: i opened on Low-Low Refueling Water Storage Tank (FWST) level in conjunction with '

a Safety Injection Signal or on a Hi-Hi Containment Pressure (Phase B) <

Containment Isolation Signal. The supply isolation valves- are 16 inch butterfly ,

valves, model number 0652, manufactured by BIF/ General Signal- Corporation (B250),.with Limitorque SMB-000 motor operators and H1BC gearboxes. Flow is controlled by a control valve on the discharge of the ND'Hx; Technical Specification 3.5.2 requires that two independent ECCS subsystems be operable in Mode 1, Power Operation, Mode 2, Startup, and Mode 3, Hot Standby.

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An ECCS subsystem is comprised of one centrifugal charging pump, one safety-injection pump, one ND pump, one ND Hx, and a flow path capable of taking suction from the FWST on an actuation signal.and automatically transferring suction to the~ Containment sump on Low-Low FWST level ^. With one ECCS subsystem  ;

inoperable, an operable status is to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or cooldown to 1 at least Mode 3 within the next six hours, and to Mode 4, Hot Shutdown, within -i the following six hours.

EVENT DESCRIPTION On October 21, 1988, PIR 0-C88-0314 was issued on valves 1(2)RN-148A, NS Heat i Exchanger 1(2)A Outlet Isolation Valves, for failure-to open under high-differential pressure during alignment for the Nuclear Service Water [EIIS:BI]

(RN) System Train A flow balance. Design Engineering (DE) used the results from Design Study CNDS-0078 (initiated in April 1988, to evaluate-seat leakage problems of BIF valves) to conclude that age hardening of seat material to a ,

mean durometer measurement of 80 Shore A is being experienced at Catawba. The manufacturer's (BIF) original specification used a 65-70 Shore A measurement in their sizing calculations. DE recommended that the actuator's open torque switch be reset to the maximum allowable position for all BIF Butterfly valves that are required to open to satisfy their safety function. There were-56 '

valves involved in this recommendation which included both the KC System, RN-System, and various ventilation systems of Units 1 andL2. By July 26, 1989, all open torque switches had been adjusted and the as-left torque switch settings for each valve was sent to DE for a final-review. j 1

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An evaluation of the final torque switch settings was documented by DE in 3 Certification of Engineering Calculations CNC-1205.02-00-0006. This evaluation-1 found that three of the actuators still may not produce the opening torque 4 required for the maximum seat hardening condition. . Valve 1KC-81B, KC to ND Hx.

7 IB Supply Isolation, and 2KC-56A, KC to ND Hx 2A Supply Isolation. were H considered to be potentially inoperable due to age hardening of the elastomeric

? seat material. The seat had been replaced in valve 2KC-81B, KC to ND Hx 2B

@ Supply Isolation, in May 1989,'under repair Work Request 42714 0PS and

significant age hardening of the seat material is not expected to have occurred l within this time. Therefore, 2KC-81B was considered to be conditionally operable. The actuator on valve 1KC-56A,'KC to ND Hx 1A supply Isolation, has a j greater output torque for its maximum torque switch setting and would not be

. restricted from opening by the age hardened seat condition. On December 13,  ;

1989, DE issued PIR 0-C89-0376 with recommendations to install a torque switch i

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bypass in the open direction on all three valves. The bypass will allow-full i

motor stall capability during unseating which will be sufficient 10 open the-  !

• valves in the hardened seat condition. Valves IKC-81B and 2KC-56A required that ,

7 the bypass be installed to eliminate operability concerns. Valve 2KC-81B would  !

require the. bypass to be installed prior to the end of the Unit 2 E003 refueling. l outage to maintain its operability. )

On December 13, 1989, at 0730 hours0.00845 days <br />0.203 hours <br />0.00121 weeks <br />2.77765e-4 months <br />, with Units 1 and 2 in Mode 1, Compliance L issued a Technical Specification Operability Notification identifying valves I L 1KC-81B and 2KC-56A as inoperable taking the most conservative approach for L

plant status. At that time, both Unit 1 and 2 entered the Technical'  !

Specification Action Statement for one inoperable train of ECCS due to a

? potentially inoperable ND Hx. Variation Notice CE-2698 (Work Request 3436 NSM)

r. and Variation Notice CE-2699 (Work Request 3437 NSM) were issued for valves j l 1KC-81B and 2KC-56A, respectively, to install the torque switch bypass. The ,

l bypasses were installed, a functional stroke-test was-performed, and Units 1 and  ;

2 exited their action statements on December 14, 1989, at 0244 hours0.00282 days <br />0.0678 hours <br />4.034392e-4 weeks <br />9.2842e-5 months <br /> and 0325 i hours, respectively.

CONCLUSION  ;

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This event has been classified as a manufacturer's functional design deficiency  !

due to an inadequate estimate of the degree of age hardening in the elastomeric seat material. The elastomeric seat material used in BIF butterfly valves is  !

subject to age-induced hardening. This tends to increase the friction force to i be overcome in unseating the valve. The increased friction was not properly '

accounted for in the manufacturer's actuator sizing calculations. ,

Review of the performance history for valve 1KC-81B revealed no occurrences (since startup) of failure to open due to mechanical binding. The performance history for valve 2KC-56A indicated that a bad wire in the motor control center resulted in a failure to open in Decmeber 1985, and that a limit switch repair was required after a stroke time test failure in October 1988. These repairs were performed under Work Request 4518 PRF and 6620 PRF, respectively. ,

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! The stroke response time for valves 1KC-818 and 2KC-56A have been trended since

$ 1987 within the ASME Section XI Valve Performance Testing Program. Both va1ves j have consistently met the 60 second response time requirement with most responses within 56-57 seconds. The trend gives no indication that the valves response time has increased with age. However. review of this data with Design Engineering concluded that thir , formation was not conclusive for determining the valves operability in rela m t to the age-hardened seat condition.

For a butterfly valve, an age-hardened seat condition would have the greatest affect on disc movement during the first few degrees of disc rotation. After the first few degrees, the age-hardened seat would have little to no affect on disc movement. A butterfly valve's performance would continue as expected _until l- the point where an age-hardened seat condition would restrict disc movement-

completely. At best, there may be only a slight increase in the response trend E over time.

From the performance history and the response time trend, it can be concluded that valve 1KC-81B and 2KC-56A were in good working condition prior to their operability determination on December 13, 1989, and that there is a high probability that the valves would have opened in response to a safety signal.

This conclusion is supported by the performance history in that no failures of o i the valves have occurred due to binding, and by the resp.nse' time trend in that a total of 10 and 11 response tests (1KC-81B and 2KC-56A, respectively) have been performed since 1987 without a failure related to seat binding. The +

issuance of the inoperability notification was a conservative approach in identifying and correcting a potentially inoperable component, albeit one proven i to be reliable by cast performance and testing.

A review of the Nuclear Maintenance Database for the past 24 months identified five BIF butterfly valves that have failed to open on-demand. Of these, only valves 1 and 2 RN-148A previously identified by PIR 0-C88-0314' failed to open due to seat binding within the seat material. A review of the NPRDS Database -

did not reveal any failures to open of BIF butterfly valves due to disc binding within the elastomeric seat material. The Operating Experience Program Database does not identify any similar occurrences within the past 24 months.

This incident is similar to other events, at Duke Power and other plants, involving valve operability under maximum design consideration conditions. This concern is the subject of NRC IE Bulletin 85-03 and Generic Letter 89-10.

Catawba Nuclear Station has established a comprehensive, on going program to ensure valves will function properly under all anticipated conditions. The thoroughness with which valve operability concerns are addressed, including DE's review of the as-left torque switch settings, contributed to the identification and prompt resolution of the current event.

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A'similar event occurred on March 14, 1988, when valve 2CA-62A, CA Pump 2A 1 Discharge to S/G 2A Isolation Valve, failed to close under high pressure drop

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conditions uni.il the valve actuator was adjusted to the maximum torque switch setting allowable. This failure was contributed to a valve factor greater than that used by the valve manufacturer in sizing the valve actuator. .The failure b of valves'to respond due to inappropriate actuator sizing by the manufacturer is L. determined to be a recurring event.

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b This event is considered to be 10CFR21 reportable. These valve failures are L

reportable to NPRDS. These valves are 16 inch butterfly valves manufactured by BIF/ General Signal Corporation (B250), model number 0652.

CORRECTIVE ACTION SUBSEQUENT

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1) Design Engineering issued PIR 0-C89-0376 identifying the effect of age-induced hardening of the seat material on the operability of o valves 1KC-81B, 2KC-56A, and 2KC-818.

2) Design Engineering issued an operability notification identifying-valves 1KC-81B and 2KC-56A as inoperable.

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3) 10CFR 50.59 Evaluation for PIR 0-C89-0376 identified 2KC-81B as conditionally operable through the end of the Unit 2 E003 refueling outage.

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l. 4) Valves 1KC-81B and 2KC-56A were modified with an open torque switch I' bypass under Variation Notice CE-2698 (3436 NSM) and CE-2699 (3437.

NSM),respectively.

! PLANNED

1) Valve 2KC-81B will be modified with an open torque switch bypass.

SAFETY ANALYSIS Two independent ECCS subsystems ensure that sufficient emergency core cooling f capability will be available assuming the loss of one subsystem through any l single failure consideration. Either subsystem operating in conjunction with l the accumulators is capable of supplying sufficient core cooling to limit the i peak cladding temperature within acceptable limits in the event of a design

!. basis loss of coolant accident (LOCA). In addition, each ECCS subsystem l- provides long term core cooling capability in the recirculation mode during the accident recovery period.

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L,[ The initial response of the ECCS to a LOCA is to provide makeup water to the NC

% System. Makeup water is provided from the FWST to the NC System via the

? charging pumps, safety injection pumps, and the ND pumps. The actual flow path p changes over time and is dependent on NC System pressure. The ND makeup flow i would not be restricted by the KC to ND Hx supply. isolation valve's failure to open. Heat removal by the ND Hxs is not required during the initial response of the ECCS to a Safety Injection Signal. Therefore, the initial response of the p ECCS to a LOCA was not impaired.

1 After the FWST is depleted, the ECCS recirculates c wling water from the

? Containment Sumps to the NC System. During recirculation, the ND Hxs and the KC System provide a heat sink for removal of decay heat from the NC System makeup.

. The opposite train of ND would have been capable of providing adequate decay heat removal if valve 1KC-81B or 2KC-56A failed to open to supply KC flow to i

their associated ND Hx.

i Control Room Operators have multiple indications of the status of the KC valves u and KC flow through the ND Hxs. Open/ Closed position of 1KC-81B and 2KC-56A is

+ indicated on-the main control board. The Engineered. Safeguards (ESF) Monitor

[! Panels 1(2) MD-4 in the Control Room include indication of the position of. these L. valves ard serve to quickly indicate abnormal alignment of equipment receiving h an ESF signal. Also, an annunciator, "ND 'Hx A(B) KC Outlet Low Flow" actuates L. if low flow exists and the respective ND pump is running. The second step of EP/1(2)/A/5000/1C3, " Transfer to Cold Leg Recirculation", directs the Operators -

to " Ensure KC flow to at least one ND heat exchanger". Operator action to'open L either 1KC-81B or 2KC-56A, if needed, could be taken locally at the valves which i are located on elevation 577 feet in the Auxiliary Building just outside the '

rooms containing the ND Hxs (as is 2KC-81B; IKC-56A.is located just inside the entrance of the Unit 1 Mechanical Penetration Room on elevation 577 feet.) It i is estimated that an Operator could locally open these valves, if needed, within 15 minutes after they were recognized as closed.

During extended operation in recirculation, the ND System has.the capability of supplying residual spray for Containment cooling to support Containment spray.

The NS System consists of two separate, equal capacity trains with each train capable of limiting Containment pressure below design restrictions during a LOCA. The ND System residual spray headers are provided to supplement NS System cooling; credit is taken for ND System spray AFTER ice bed depletion IF one train of NS spray is unavailable. Failure.of a KC to ND Hx Supply Isolation valve to open would have prevented the associated ND train from supplying cooled residual spray. One ND train has the capacity to provide sufficient residual spray as well as adequate core cooling flow via one centrifugal charging and one safety injection pump (ND injection into the NC System is manually isolated prior to swapping ND train to residual spray).

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i These valves, had consistently passed their stroke time test with no indication of an adverse trend resulting from the age-hardening process. However, c

calculations indicated that the potential' existed for age-hardening to affect l: the valves ability to open under the maximum differential pressure condtion. ,

f; Therefore, the valve was adjusted for maximum torque to further assure l l: operability. It is concluded that the health and safety of'the public were not )

l:- adversely affected.

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