05000423/LER-1996-035-01, :on 960920,motor Operated Valve Performance Found Outside Design Basis of Plant.Caused by Inadequate Design & Verification of Actuator.Mov Calculations Will Be Revised

:on 960920,motor Operated Valve Performance Found Outside Design Basis of Plant.Caused by Inadequate Design & Verification of Actuator.Mov Calculations Will Be Revised

ML20129A187
Person / Time
Site:	Millstone
Issue date:	10/18/1996
From:	Peschel J NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20129A114	List: 05000423/LER-1996-035, Forwards LER 96-035-00.MOV Calculations Will Be Revised Based on Bounding Application Factor/Pullout Efficiency or MOV Specific Performance Characteristics Prior to start-up from Current Outage 05000423/LER-1996-035-01, :on 960920,motor Operated Valve Performance Found Outside Design Basis of Plant.Caused by Inadequate Design & Verification of Actuator.Mov Calculations Will Be Revised
References
LER-96-035-01, LER-96-35-1, NUDOCS 9610220053
Download: ML20129A187 (4)

LER-1996-035, on 960920,motor Operated Valve Performance Found Outside Design Basis of Plant.Caused by Inadequate Design & Verification of Actuator.Mov Calculations Will Be Revised

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(t)(2)
4231996035R01 - NRC Website

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROYED BY OMB NO. 3160-0104 EXPIRES 04/30/98 (4-9 51 EoRUAr C ICT ON A UEST 50 RS E A TED S

t'a"'?o'%51# "*!BKe 'MJ=s" J"?f!'a ^".82 LICENSEE EVENT REPORT (LER) 15;t^"# 'Mia"^'M '&"i#J,."s'# '"'s,3"'s" R

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digits / characters for each block)

FACILITY NAME (1)

DOCKET NUMBER 123 PAGE (3)

Millstone Nuclear Power Station Unit 3 05000423 1 of 4 TITLE 44)

Motor Operated Valve Performance Outside the Design Basis of the Plant EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FActUTY NAME DOCKET NUMBER NUMBER 09 20 96 96 035 00 10 18 96 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS CF 10 CFR 5: (Check one or more) (11)

ODE N 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i)

X so 73(a)(2)(ii) 50.73(a)<2)ixi LEVEL (10) 20.2203(a)(2)(i) 20 NO3(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER g,

20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v)

Specify in Abstract below or in NRC Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii) r LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (include Area Codel J.M. Peschel, MP3 Nuclear Licensing Manager (800)437 5840 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER HEPORTABLE

CAUSE

SYSTEM COMPONENT MANUF ACTURER REPORTABLE To NPRDs To NPROS i

I SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR SUBMISSION f NO YES (if yes, complete EXPECTED SUBMISSION DATE).

l ABSTRACT (Limit to 1400 spaces, Le., approximately 15 single-spaced typewritten lines) (16) i j

On September 20,1996, while the plant was in Cold Shutdown Mode, an evaluation was performed on all Motor Operated Valves (MOVs) within the scope of Generic Letter 89-10 to determine if they would have stroked under design basis conditions based on the use of Limitorque's Pullout Efficiency and an application factor (APFR) of 0.9.

The review identified 27 MOVs that potentially may not have stroked fully under design basis conditions using the j

most conservative design input assumptions available. This condition is being reported pursuant to 10 CFR 50.73 (t)(2)(li) as a condition outside the design basis of the plant.

I j

This evaluation was predicated by NRC issuance of Information Notice 96-48, " Motor Operated Valve Performance issues" dated August 21,1996. The IN questioned the use of the vendor supplied data for MOV thrust calculations in determining the actuator capability to stroke against their respective design basis conditions. This evaluation idsntified 27 MOVs that may not be capable of stroking apa!rJt their de::gn basis differential pressure at this newly ezsumed actuator efficiency. The identified MOVs have oeen determined to be inoperable.

9610220053 961018

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PDR ADOCK 05000423 4

S PDit NRC FORM 366 (4-95) 4 i

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.,U.S. NUCLEAR REoVLAToRY Commission

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4 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION l

rACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVislON Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 4 00 96 035 TEXT (11more space is required, use additional copies of NRC Form 366A) (17) l.

Description of Fy.e,n1 l

On Septs mber 20,1996, while the plant was in Cold Shutdown Mode, an evaluation was performed on all Motor Operated Valves (MOVs) within the scope of Generic Letter (GL) 89-10 to determine if they would have stroked under d* sign basis conditions based on the use of Limitorque's Pullout Efficiency and application factor (APFR) of 0.9. The review identified 27 MOVs that potentially may not have stroked fully under design basis conditions using the most conservative design input assumptions available. The following MOVs were identified as lacking capability and dedared inn,nerable:

3CHS*LCV112D 3CHS*LCV112E 3CHS*MV8110 3CHS*MV8111B 3CHS*MV8468A 3CHS*MV8468B 3CHS*MV8511 A 3CHS*MV8511B 3CHS*MV8512A 3CHS*MV85128 3FWA*MOV35B 3 MSS *MOV18A 3 MSS *MOV18B 3 MSS *MOV18C 3 MSS *MOV18D 3 MSS *MOV74A 3 MSS *MOV74B 3 MSS *MOV74C 3 MSS *MOV74D 3RHS*MV8701A 4

3RHS*MV8701C 3RHS*MV87028 3RHS*MV8702C 3SIL*MV8808A 3SIL*MV8808B 3SIL*MV8808C 3SIL'MV8808D i

information Notice (IN) 96-48, " Motor Operated Valve Performance issues", identified potential non-conservative assumptions in adaptations to the Limitorque standard sizing equation for determining actuator thrust and torque capability under design basis conditions. Limitorque Technical Update 93-03 allowed the use of an APFR 1.0 if the voltage was below 90% of the rated nameplate voltage. Additionally, Limitorque allowed the use of running efficiency i

vice pullout efficiency for an AC motor without throttling capability for the close stroke i

11.

Cause of Event

The cause of the event is inadequate design, and verification of actuator and motor performance characteristics. The i

design of the identified MOVs incorporated the use of running efficiency and/or motor application factor of 1.0.

Incorporating pullout efficiency and/or a motor application factor of 0.9 reduced the actuator capability to a point less than that required to stroke the MOV(s) against their respective design basis conditions.

Ill. Analysis of Event This evaluation was predicated by NRC issuance of IN 96-48, " Motor Operated Valve Performance issues

• dated i

August 21,1996. The IN questioned the use of the vendor supplied data for MOV thrust calculations in determining the actuator capability to stroke against their respective design basis conditions. The IN stated that the use of a 1.0 spplication factor in conjunction with runring efficiency may not bound the motor actuator performance characteristics b sed on limited industry testing. Tbceafore, preliminary calculations were performed using an application factor of 0.9 and pullout efficiency for all GL 99-10 MOVs in the plant. This evaluation identified 27 MOVs that may not be capable of stroking against their design basis differential pressure at this newly assumed actuator efficiency. The 4

identified MOVs have been determined to be inoperable.

NRC FORM 306A 14 95)

l 4

N:',C FOIM 366A U.S. NUCLEAR REGULATORY CoMMissd LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3) i YEAR SEQUENTIAL REVISION i

Millstone Nucke Power Station Unit 3 05000423 NUMBER NUMBER 3 of 4 035 00 96 TEXT lif more space is required, use additional copies of NRC Form 366A) t17) l i

4 l

?

l

IV. Corrective Action

i The MOV actuator efficiency and motor capability for each valve will be determined by performing on site testing, e

or, by utilizing bounding performance characteristics (application factor and efficiencies) from industry testing prior to start-up from the current outage.

The MOV calculations will be revised based on bounding application factor / pullout efficiency or MOV specific e

2 performance charactenstics prior to start-up from the current outage.

The actuator torque switches will be reset based on the newly calculated setpo;r.ts prior to start-up from the e

current outage.

MOVs will be modified as required to ensure adequate capability of the MOV to stroke during their credited design e

basis event (s). These modifications will be completed prior to start-up from the current outage.

V.

Additional Information

l Not Applicable a

Similar Events

LER 96-019-00:

Reactor Coolant System Power Operated Relief Valve Block Valves inoperable due to i

Potential Structural Desian Deficiency i

On June 27,1996, at 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br /> with the plant shutdown in Mode 5 the Reactor Coolant System (RCS) Power Operated Relief Valves (PORV) Block Valves (3RCS*MV8000A/B) were determined to be unable to perform their intended safety functions to close and reopen under design basis accident conditions. Tests performed at Kalsi Engineering Inc.

(KEI) provided evidence showing the valves would require greater thrust to close than had been previously calculated, and damage to the valve during attempted closure under design basis conditions could prevent reopening. An immediate notification was made at 1454 hours0.0168 days <br />0.404 hours <br />0.0024 weeks <br />5.53247e-4 months <br /> on June 27,1996, pursuant to 10CFR50.72(b)(1)(ii)(B) for a condition outside the design basis of the plant. The cause of this event appears to be a structural design deficiency.

Failure of the valves to perform their required opening or closing function during design basis events could result in difficulty controlling Reactor Coolant System (RCS) pressure and inventory, thereby potentially increasing the severity of an accident, j

Further full scale testing was performed by Kalsi Engineering using representative components to determine the full extent of the problem. Modifications to the valve internals, or replacement of the valves will be performed.

j LER 94-004-00 Feedwater Isolation Valves Potentially f nocerable as a Historical Condition 4

fCC FORM 368A (4-95) d

NRC FORM 363A U.S. NUCLEAR REGULATORY COMMISSION (4 95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 4 of 4 96 035 00 TEXT tilmore space is required, use additional copies of NRC form 366A) 817)

Manufacturer Data Ells System Codes Auxiliary Feedwater-BA Residual Heat Removal-BP ChrmiCal and Volume Control-CB r

Main Steam - SB Ells Eauipment Codes Motor Operated Valve - 20 i

l F

I MRC FORM 366A 14-95)

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