ML20046C216

LER 93-017-00:on 930701,possible Inoperability of Power Operated Relief Valve Blocking Valves Caused by Original Design Basis.Changed Designs & Performed tests.W/930803 Ltr

ML20046C216
Person / Time
Site:	Millstone
Issue date:	08/03/1993
From:	Nichols D, Scace S NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-017, LER-93-17, MP-93-616, NUDOCS 9308100003
Download: ML20046C216 (5)

Text

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NORTHEAST UTILITIES o===Saa S"* Sa conaa"out 9 we., n. eYc RO. BOX 270 i

' 2 ZL*d* ",7d*4 HARTFORD. CONNECTICUT 06141-0270 Norewast Nudem Enwgy Company (203)665 -5000 1 August 3, 1993 MP-93-616 Re: 10CFR50.73 s

U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Reference:

Facility Operating License No. DPR-65 Docket No. 50-336 Licensee Event Report 93-017 Gentlemen:

This letter forwards Ucensee Event Report 93-017. This report is submitted for information only.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY FOR: Stephen E. Scace Vice President - Millstone Station

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BY: hn S. K enan 8 nit 2 Director r SES/ din

Attachment:

LER 93-017 cc: T.T. Martin, Region l Administrator P. D. Swetland, Senior Resident inspector, Millstone Unit Nos.1, 2 and 3 G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 090097 ei l

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• LICENSEE EVENT REPORT (LER) ""'* "*"""$ET""EIE'S**(P nEPORTS NAGE 30) US LEAR REGLAATORY COMMISSON, WASHINGTON, DC 20$55. AND TO THE PAPERWORK REDUCTON PROK CT (3150-0104). OFT 1CE OF MANAGEMENT AND BUDGET, WASHINGTON. DC 20503.

FafluTY NAME 0) DOCKET NUMDER (2) PAGE f 31 I Millstone Nuclear Power Station Unit 2 0]5 l0 l0 l0 l3 l3 l6 1 oFl 0l 4 IITLE (4)

Possible Inoperability of Power Operated Relief Valve Blocking Valves EVENT DATE (5) LIR NUMBER f8) REPORT DATE (7) OTHER FACIUTIES INVOLVED fB)

MONTH DAY YEAR YEAR N MONTH DAY YEAR FACluTY NAMES

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of slo l 0101 l l 0l7 0l1 9 3 9 l3 0l1l7 0l 0 0l8 0l3 9j 3 , y , ,

OPERATING THl3 REPORT iS BEING SUDMITTID PURSUANT TO THE REOUIREMENTS OF 10 CFR I (Check one or more of the tonowing)(11) i N 20.402(b) 20 402(c) 60.73(a)2)$v) 73.71(b) l PCP#liR 20 40S(a)(1)(1) 50.36(c)(1) 60.73(a)G)M

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NAME TELEPHONE NUMBER AREA CODE D. L.Nichols, Associate Engineer 2l013 414171 -11171911 COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN TFRS REPORT (13)

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[ NO l l l l raSTRACT (Umst to 1400 apaces. t.o., approsomately theen eingle-space typewrftten Ines) (16)

On July 1,1993, with the reactor in Mode 1, at 100% power, discussions with NRC Representatives, relative to implementation of NRC Generic Letter 89-10, identified that manufacturer's design valve factors, used in

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determination of operability of the Power Operated Relief Valve (PORV) Blocking Valves,2-RC-403 and 2-RC-405 prior to Operating Cycle 12, may not have been sufficiently conservative. Preliminary results from NRC and industry test programs indicate some gate valves require more force to close than hr_J been assumed by the manufacturer's valve factor. Use of the manufacturer's design valve factors led to a conclusion that these valves were operable when they may not have been able to close to isolate a stuck open PORV at design basis conditions. The possible effects of using a non-conservative valve factor are tempered by other conservatisms in the calculations. Design changes and testing performed during the nefueling prior to Operating Cycle 12 provide assurance that the valves are capable of fulfilling all design basis functions during Operating Cycle 12 and subsequent cycles. This report is filed forinformation only.

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4 C orm 306A U.S. P'UCLEAR REGUIATORY COMMISSION APPROVE OM NO 0104 ESTIMATED BUHDEN PER RESPONSE TO COMPLY WITH THIS LICENSEE EVENT REPORT (LER) INI- LLE N

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DOCKET NUMBER Q) (ER NUMBER im PAGE (3)

YEAR E Millstone Nuclaar Power Station Unit 2 ~ -

ten a m,. .p.e. m r.qar.o u.. eddlianal NRC Fom* 366A's) {I f) ols o l0 o l3 l3 ls 9l3 0l1l7 0l0 0l2 OF 0l4 L D9sCIjpliRTLQMyant On July 1,1993, with the reactor in Mode 1, at 100% power, discussions with NRC Representatives relative to implementation of NRC Generic Letter 89-10 identified that manufacturer's design valve factors, used in determination of operability of the Power Operated Relief Valve (PORV) Blocking Valves,2-RC-403 and 2-RC-405 prior to Operating Cycle 12, may not have been sufficiently conservative. Preliminary results from NRC and industry test programs indicate some gate valves require more force to close than had been assumed by the manufacturer's valve factor. Use of the manufacturer's design valve factors led to a conclusion that these valves were operable when they may not have been able to close to isolate a stuck open PORV at design basis conditions. The possible effects of using a non-conservative valve factor are tempered by other conservatisms in the calculations. Design changes and testing performed during the refueling prior to Operating Cycle 12 provide assurance that the valves are capable of fulfilling all design basis functions during Operating Cycle 12 and subsequent cycles.

fl. Cause of Event The root cause is attributed to use of valve manufacturer recommended valve factor (original plant design basis) in the analysis of motor operated valve operability issues. Preliminary NRC and industry test data now indicates that some gate valves may have a higher valve factor than that provided by the manufacturer.

Valves 2-RC-403 and 2-RC-405 are designed to close against a differential pressure of 2250 psid. Prior to Operating Cycle 12 the valves were maintained and adjusted using the accepted and evolving engineering methodology of that time. Design basis reviews, conducted in response to NRC Generic Letter 89-10, were initially based on the manufacturer's design basis valve factor (for closing direction) of 0.3.

During the ongoing efforts to improve operational readiness of motor operated valves, actuator sizing calculation methodologies have been changed to include several additional factors. These factors include, revised estimates of test equipment accuracy, reduced (more conservative) under-voltage values, revised valve seat areas, consideration of piston effects, in-depth analysis of pressure, temperature, and flow characteristics, and replacement of estimated or assumed parameters with measured values provided by state-of-the-art diagnostic equipment. Preliminary results from NRC and industry sponsored testing indicates that some gate valves require more force to close than had been assumed by the manufacturer's valve factor (such test results are specific to valve c:esign and test conditions). In addition, industry experience and vendor provided information indicate that test equipment used to establish torque switch trip settings prior to Operating Cycle 11 may not have provided the anticipated accuracy. Thus, calculations supporting torque switch settings and operanility determinations which were based on the manufacturer's valve factor may not have been sufficienby conservative. Because these calculations were used as the basis for adjustment of valve actuator control switches, these valves may have been unable to close against 2250 psid. Because of the uncertainties and questionatile diagnostic equipment accuracies inherent in work performed prior to Operating Cycle 12 a more concrete determination of operability is not practical.

lif. AnalysiSSEYRDt This event is being reported for information only. Since indisputable evidence cannot be gathered to fully support past operability of the subject valves, this report is provided to discuss the uncertainties of work performed prior to Operating Cycle 12. At no time was plant safety at risk due to the potential inability of the PORV blocking valves to close against 2250 psid because there is sufficient margin available in the valve and actuator design to provide assurance that the valves would have been capable of isolating a stuck open PORV as pressure decreased.

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ESTIMAftD BURDEN PER RESPONSE TO COMPLv Wim THIS

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DOCKET NUMBER g)

LER NUMMR (6) PAGE (3) vtAR emuEum. g Millstone Nuclear Power Station

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ols jo lo lo l3 l3 le 9l3 0l1l7 0l0 0l3 OF 0l4 itxT of more space is raped, use adsanal NRC Form 36SA's) (m This margin is assured by:

Ana!ysis of anticioated valve differential oressure. The design basis differential pressure is conservative.

Analysis shows that pressure upstream of the block valves will be 1400 psid when isolating a stuck open PORV 60 seconds after it opened at full power (assumes a thermal margin / low pressure trip at 56 seconds).

Without a trip the differential pressure would be 1880 psid. If the operator is assumed to wait 10 minutes to take action from the Control Room (a required assumption in the performance of safety evaluations and operability determinations) the differential pressure would decrease to 800 psid. Any reduction in pressure when the valve is called on to operate increases the available margins and the probability that the valve will perform as desired.

Review of the control circuitrv. If the valves reach their control switch setpoint prior to full closure, the control circuit, which has a maintain-position control switch, will allow the motor to re-energize when -

pressure decreased and the spring pack relaxes. The motor actuators have the capability to restart with sufficient thermal margin to complete full closure, {

j Use of coDEervatiys_under-voltage _ Calc _ulations. Design basis calculations are based on locked rotor amperage. Since actual starting current is lower, actual applied voltage would be higher and provide more power to the actuator.

Exoerience with sirnitar valves. The valve vendor (VELAN) has reported that these 2-1/2 inch valves have performed well under PORV isolation conditions in the 1982 EPRI-Marshall tests. Although some industry data indicates gate valves subject to line break conditions may have valve factors as high as 0.6, these valves are not subject to line break conditions because of the back pressure created by the downstream PORV, discharge piping and quench tank. {

In summary, although engineering assessment, using current methodologies, cannot conclusively confirm that these valves would have functioned properly under design basis conditions prior to Operating Cycle 12, valve mis-operation would not have had a significant impact on safety, since valve closure would eventually occur as pressure decreased.

IV. Corrective Actior; During the refueling outage prior to Operating Cycle 11 and Operating Cycle 12, actuator spring pack and gear changes were accomplished to increase available thrust. These actions assure that the valves are currently operable at design basis conditions and have up to 50% margin to account for potential i uncertainties in valve design factors. Engineering studies are being performed to determine an appropriate '

course of action in the event that industry sponsored testing identifies that more than 50% uncertainty margin is wananted I

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+ 's' s; NRC Fofm 386A g-es) U.S. NUCl. EAR REOULATORY COMMISSION APPROYED OMB NO,3150-0104 E;JTtES: 4/30E2 ESTIMATED BURDEN PER RESPONSE 10 COMPLY WITH THIS

'* LICENSEE EVENT REPORT (LER)  %%5,","C'S"57,%g8,hFORW TEXT CONTINUATION 7 s

. AND REPORTS MANAGEMENT BRANCH (P-830), U.S. HUCLE.AR -

REGULATORY COMMISSeON, WASHMGTON. DC 20566. M4D TO THE PAPERWORK REDUCTION PRCAECT (3150-0104). OFFICE OF -

MANAGEMENT NO BUDGET WASHINGTON. fT 9503.

FAQuTY NAME (1) DOCKET NUMBER Q) 1.ER NUMBER (a PAGE (3)

YEAR Millstone Nuclear Power Station Unit 2 - ~ ~

TEXT pf more space is reqwred, use addsonal NRC Form 3664's) (17) ol5 l0 l0 l0 l3'l3 le 9l3 0l1l7 0l0 0l4 OF 0(4

' V. AddttLQDalInformatior)

Actuator Data Manufacturer - Umitorque Corp.

Model - SMB-000 Spdng Pack No. 0101-093 Overall Ratio - 68.00:1 Ells No. - AB-84-L200 Valve Data Manufacturer - Velan i Drawing No. - P2-0634-N-16 Nominal Size 1/2 inch a '

Pressure Rating - 2500 psi .

Type - Gate i Design Valve Factor - 0.3 Ells No. - AB-20-V085  ;

Similar Ucensee Event Reports - None 1

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