05000336/LER-1987-009

LER 87-009-02:on 870902,unit Experienced Automatic Reactor Trip Due to Low Steam Generator Level.Caused by Valve Stem Slightly Loose from Valve Plug.Damaged Stem/Plug Assembly replaced.W/900718 Ltr

ML20055H012
Person / Time
Site:	Millstone
Issue date:	07/18/1990
From:	Bergin J, Scace S NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-87-009-01, MP-90-717, NUDOCS 9007250049
Download: ML20055H012 (4)

LER-2087-009,

Event date:
Report date:
3362087009R00 - NRC Website

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< July'18,.1990 MP-90-717.

Re: 10CFR50.73(a)(2)(iv) ,

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U.S. Nuclear Regulatory Commission Document Control Desk o Washington, 'D.C; '20555 P-

Reference:

Facility Operating License No' DPR-65 Docket No. 50-336 Licensee Event Report 87-009-02  ;

i Gent l emen:

l1This11etter- forwards Licensee Event Report 87-009-02 required' to be submitted pursuant to paragraph 50.73(a)(2)(iv). l 4

Very truly yours, 1

NORTHEAST NUCLEAR ENERGY COMPANY l

FOR: Stephen E. Scace j Director, Millstone Station l ]

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L -BY:

k n P. Stetz Millstone Unit 1 Director SES/JMB:mo l

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Attachment:

LER 87-009-02 1,

L cc: T. T. Martin, Recion 1 Administrator l: W. -J. Raymond,' Senior Resident inspector, Millstone Unit Nos.1, 2 and 3 G. S. Vissing, NRC Project Mar.ager, Millstone Unit No. 2 L

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.NRC Form 366 U.S. NUCLE AR REGULATORY COMMISSION APPROVED OMB NO. 3160-0104 (6-89)* EXPERE S: 4/30/02

'a Estimated buroen per respons2 to comply witn this Information cohection teovest: 50.0 hrs Forward LICENSEE EVENT REPORT (LER) ' Me"cUtEflNeNint B .NE*2 ')"*uIS 30 Uu'cioar Regulatory Commission, Washington. DC 20555. and to f the Paperwork Reduction Protect (3150-0104). Office of Management and Buopet Washington. DC 20503.

FACILITY NAME (1) DOCF.E I NUMBEH (2) 8%M Millstone Nuclear Power Station Unit 2 ol sl ol ol ol3 l3 ls i lod 0l3 TITLE (4)

Reactor Trip on Low #1 Steam Generator Level EVENT DATE (51 LFA NUMPF A #61 AFPORT D ATE t71 OTHFA F ACrtl TIFS INVOLVED (Si MONTF- DAY YEAR YEAR MONTH DAY YEAR FAOldTY NAMES olslofofof I I 0 9 0l2 8 7 8 l7 0l0l9 0l 2 0l7j1l8 9l0 01610l o1 ol l l OPERATINo THIS REPORT IS BEINo SUBMITTED PURSUANT TO THE AEOtrREMENTS OF 10 CFR $ ICneck one or more of the followingH11) 20 402(b) _

20 402(cl J 50.73(a)(2)(ev) _

73.7 t (b) i po a 20.405(a)(1)(0 50.36(C)(1) 50.73(a)(2)(v) 73.71(c)

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NO l l l AssTRACT (Limit to 1400 spaces. e e . amoroximateiy titteen singie-space typewntion nresi (i6; While operating at 91re power on September 2,1987 at 2015 the unit experienced an automatic reactor trip due to low steam generator level in the #1 steam generator. The operation's staff was performing a routine reduction in power level from 100re to 90Fe. Dunng this evolution the secondary plant operator observed the feed flow to the #1 steam generator less than the steam flow. The operator took manual control of the feedwater regulating valve, 2-FW-51 A, and the feedwater regulating bypass valve, 2-FW-41 A, in an attempt to l restore level in the #1 steam generator. Level continued to decrease in the #1 steam generator and the unit tnpped. Operations responded to the trip by performing EOP 2525, " Standard Post Trip Actions" and EOP i 2526 " Reactor Trip Recovery" No other systems were affected and the unit was placed in a stable condition.

The #1 feedwater regulating valve, 2-FW-51 A, was disassembled and repaired. Dunng the disassembly it was l discovered that the stem had separated from the plug. This event is being reported pursuant to the requirements of paragraph 5L.73(a)(2)(iv) due to the automatic reactor trip on low steam generator level.

Similar LER's: None.

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1. Descrintion of Event While operating at 91rc power on September 2,1987 at 2015 the unit experienced an automatic reactor trip due to low steam generator level in the #1 steam generator. The operation's staff was performing a-routine reduction in power level from 100rc to 90re. The power reduction was in response to fouling of-the B condenser waterbox, requiring it's removal from service. During this evolution the secondary plant operator observed the feedflow to the #1 steam generator to be less than the steam flow. The operator took manual control of the feedwater regulating valve, 2-FW-5' A and the feedwater regulating bypass valve, 2-FW-41 A, in an attempt to restore level in the #1 s'erm generator, i.evel continued to decrease '

in the #1 steam generator. Plant Equipment Cperators were dispatched to the feedwater regulating valve, ,

2-FW-51 A, to take local manual control. The Unit tripped prior to the Plant Equipment Operators i arriving at the valves. The licensed operators responded to the trip by performing EOP 2525, " Standard 5 Post Trip Actions" and EOP 2526, " Reactor Trip Recovery" No other systems were affected and the unit was placed in a stable condition.

11. Cante of Event Investigation of this event included the disassembly of the #1 feedwater regulating valve, 2-FW-51A, to l determine the cause of the lack of. response to automatic or manual level control signals. During the disassembly the valves bonnet stem and p!ug were removed from the valve body. Further investigation revealed that the valve stem was slightly loose from the valve's plug. The stem was then torqued in an -

attempt to tighten the stem to the plug. The stem then became loose enough to remove the stem from the plug by pulling it out of the plug by hand. The threads on the stem were rolled over and were not applying the required connection between the plug and the stem. The threads on the plug are also in a

- similar condition. Proper connection between the stem and the plug for this valve requires the stem to be threaded into the plug so that a tapered portion of the stem makes contact with a tapered fit on the plug. Rotation of the stem with respect to plug is resisted by the friction forces in the threads and a roll pin which is installed through the stem and plug assembly. Examination of the valve body and the stem / plug assembly did not reveal the existence of the roll pin. The Unit staff has concluded that the root cause of the transient was the loss or the lack of this roll pin in the stem plug assembly initiated the degradation of the stem / plug assembly which eventually resulted m the loss of feed water flow control and steam generator level.

l 111. Analysis of Event l This event is being reported pursuant to the requirements of paragraph 50.73(a)(2)(iv) due to the I automatic reactor trip on low steam generator level.

There were no safety consequences resulting from this reactor trip since all safety systems functioned to restore the Unit to a stable condition. Aher the reactor trip the damaged valve received a signal to close I with the feedwater regulating bypass valve receiving a signal to open. The bypass valve supplied enough

flow to recover level in the al steam generator. The introduction of the rolled pin into the steam generator has been reviewed to evaluate its affect on the integrity of the steam generator tubing. This l condition has been found acceptable due to the slignt mass of the pin involved. Since it cannot be i determined if the rolled pin was installed by the valve's manufacturer a shght potential may exist that this same condition may occur in the #2 steam generator feed water regulating valve. This condition has been determined to be acceptable based on the low probabihties of occurrence and the ability to provide an adequate heat sink to the affected steam generator via the feedwater regulating bypass valve for removal of decay heat. The Unit was shutdown for 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> 20 minutes.

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-IV. Corrective Action The damaged stem / plug assembly was replaced, the valve was assembled and tested. The Unit was then returned to service. This valve and the #2 feedwater regulating valve have been inspected during the End of Cycle 09 refuel outage, no problems of a similar nature were found in these valves. The appropriate valve preventative maintenance procedure has been revised to inspect for the proper installation of this roll pin.

.V. Additional Information

. The Main Feedwater Regulating valve, 2-FW-51 A, is a Copes - Vulcan 12" 900# angle style valve ~ l assembly with a P-200-12 operator. '

Similar LERs: None l-l l

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