ML19325C187

LER 87-009-01:on 870902,unit Experienced Automatic Reactor Trip Due to Low Steam Generator Level in Steam Generator 1. Caused by Lack of Roll Pin in Stem Plug Assembly.Damaged Stem/Plug Assembly replaced.W/891002 Ltr

ML19325C187
Person / Time
Site:	Millstone
Issue date:	10/02/1989
From:	Bergin J, Scace S NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-87-009-01, LER-87-9-1, MP-13568, NUDOCS 8910110248
Download: ML19325C187 (4)

Text

-

~

p ., ,

'a '

f . , g;g g m ,

o.n.js cm e:..s..n simi. s.sn conn.ciied i?.50

-( " .Iy EdeYmsNNttNec CNn"5 y P.O. BOX 270

[' ?!:-

{. ..' gggg,gC7 & ny Northeast Nuclear Energy Company --

HARTFORD. CONNECTICUT.06414-0270 -

.j203)665-50001 tj

(

1

-- [

October'2, 1989

.MP-13568-

. Re: 10CFR50.73(a)(2)(iv)  ;

1

~1

, 5 e

lf.S/ Nuclear Regulatory. Commission Document Control Desk ,

Washincton, D.C. 20555

Reference:

Facility Operating License No. DPR-65

. Docket No. 50-336 Licensee Event Report 87-009-01 J Gentlemen:

' This letter forwards the update Licensee Event Report 87-009-01 required to be

~

submitted pursuant to paragraph 50.73(a)(2)(iv). t i

Very truly yours,

NORTHEAST NUCLEAR ENERGY COMPANY A

f,ge r St p E. cace .

Station Superintendent Millstone Nuclear Power Station- l L

L ,.

SES/JMB:mo

Attachment:

LER 57-009-01 cc: W. T. RusseU, Recion I Administrator.

~

W. Ji Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2 and 3 G. S. Vissing. NRC Project Manager, Millstone Unit No. 2 i

1.

/( f/

. , 8910110248 891002.

L PDR ADOCK 05000336 l' -

S PDC v _ _

NRC Form 366 V S NUCLEAR REGULATORY COMMISSION APPROVLD OMB No 316 > 0104-g (6-4C9 ;

• EXPIRE S: 4/30/92 Estimated twecen per response to comoiy with Inis >

mtsmation coisemson reosest: 60.D hrs. Forwted

• - *1 LICENSEE EVENT REPORT (LER)- $ T oo'[s7/a M .# "nen Br$' U E$0) u S E'ies-Re0ulato,y Commission. Washmoton. DC 20665. arks to e' '? the Pape*wo'k Reduction Project (3t60-Ottd) office of Management and Esuccet. Washington. oC 20603.

FACluTY NAME (1) DOCPil NVMBER (2) 4GE -*

Millstone Nuclear Power Station Unit 2 of sl ol ol ol3 l3 ls 1lor 0l 3 inct .)

Reactor Trip on Low #1 Steam Generator Level EVENT DATE f 51 LER NUMBER f6s i REPORT DATF (71 OTHi!b F AC1UTIE 8 NVOLVED (8)

MONTF DAY YEAR YEAR N NM MONT H DAY YEAR FACluTY NAMES

~ ~

0 9 0l2 8 7 8 l7 0l0l9 0l 1 1l0 0l2 8l9 of si ol of of l l OPERATNo THIS REPORT 18 BENG SUBM.TTED PURSJANr to THE REOUIREMENTS of 10 CrR 3. (Cneck one ce more of tne toltowm0)I11)

E @) 20 402tti, 20.402(ci 1 60.73(a)(2)Ov) 73.7 t (D)

R _

20 406(a)(1)D) 60.3 (c)(1) 60.73(a)(2)(v) 73.71(c)

1. ion' l9l1 _

20 406(a)(1)01) 60.36(c)(2) 60.73. (a)(2)(vil) _ {T R

goe it 20 406(a)O)Oii) 60.73(a)(2)(l) 60.73(a)(2) tvill); A) Text. NRC Form 366A) .

60,73(alf2Hii) 20 406(eH1)(iv) 30.73ta)(2)(villHB) 20 406ta t(11tiv) l 60.73(a)(210lf) 60.73(a)(2) t x)

UCENSEE CONT ACT FOR THis LER (12n NAME TELEPHONE NUMBER AHEA CODI Joseph M. Bergin, Engineer Ext. 5352 2l0l3 4l4l7l-l1l7l9l1 COMPLETE ONE UNE FOR E ACH COMPONENT F A; LURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT $URygC- d CAUSE SYSTEM COMPONENT Mh k* M X SlJ LlClV) Cl 6 l 3 l5 l l l l l l l l 1 l' l i I I c I Il1 I I I SUPPLEMENTAL REPORT EXPECTED fidi MONTH DAY YEAR g

SUBM;S$lON YES of yes. comoiste EXPECTED SUBM;SSION DATE) A 06)

% No l l l

• BSTRACT u,,t to i.00 soaees, ... . approxeat., tme.n .,e-soac. ipe ntion im.si nel 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 performinE a routine reduction in power level from 100Fe to 90rc. During this evolution the secondary plant operator observed'the feed flow to the #1 steam generator less than the steam flow. The operator took manual l

control of the feedwater regulating valve, 2-FW-51B, and the feedwater regulating bypass valve, l'

2-FW-41B, in an attempt to restore level in the #1 steam generator. Level continued to decrease in the

1. 1 steam generator and the unit tripped. Operations responded to the trip by performing EOP 2525.

" Standard Post Trip Actions" and EOP 2526 " Reactor Trip Recovery" No other systems were affected and the unit was placed in a aable condition. The #1 feedwater regulating vahe, 2-FW-51B, was disassembled and repaired. During the disassembly it was discovered that the stem had separated from the l plug. This event is being reported pursuant to the requirements of paragraph 50.73(a)(2)(iv) due to the l automatic reactor trip on low steam generator level. Similar LER's: None.

l orm 366

^

y '

.j .

U.S. NUCLE AR REoVLATORY CoMMISSloN MPRoV O 60-DID4

-[ gform 366A

....,na,.mo., n., r..- .. ,.~n ini, g .

LICENSEE EVENT REPORT (LER) ["j",,m*2,7,gt,g.yosygy;.,,,Fv, ,gero,,_

TEXT CONTINUATION .no R.nors uanao.m.nt erance. ap-53ci. u.s Naciw 1 a . on Pro 3 60- 104 of o Manap.mont anc Baopet. Washmoton. DC 20s03 '

F ACUTY NAME (1) DOCKET NUMBER (2) (E A Nuverm rei PAGE13) l W

h YEAR N 1

= Millstone Nuclear Power Station "- ""

Unit 2 og, ogogog3y3g, gg7 0l 0l 9 0l1 0l 2 OF .0l 3 TEXT (tf more .nace as r.quir.c. u.e sooitional NRC Form 366A 5) (17)

1. Deerintion of Event -

While operating at 917e 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 senice. During this a l' 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-51B, and the feedwater regulating bypass valve, 2-FW-41B, in an attempt 'o restnre Ievel in the #1 steam generator, l eve 4 continued to decrease m the #1 steam generator. Plant Equipment Operators were dispatched to the feedwater regulating valve, 2-FW-51B, to take local manual control. The Unit tripped prior to the Plant Equiprpent Operators arriving at the valves. The licensed operators responded to the trip by performmg EOP 2525. " Standard Post Trip Actions" and EOP 2526,' " Reactor Trip Recovery". No other systems were affected and the unit was placed m a stable condition.

11. Cnme of Event Investigation of this event included the disassembly of the #1 feedwater regulating valve, 2-FW-51B, to determine the cause of the lack of response to automatic or manual level control signals. During the disassembly the valves bonnet stem and plug were removed from the valve body. Further investi;ation revealed that the valve stem was slightly loose fron. 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 in the loss of feed water flow control and steam generator level.

III. Annivsic of Event This event is being reported pursuant to the requirements of paragraph 50.73(a)(2)(iv) due to the 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. After the reactor trip the damaged valve received a signal to close with the feedwater regulating bypass valve receiving a signal to open. The bypass valve supplied enough flow to recover level in the #1 steam generator. The introduction of the rolled pin into the steam generator has been reviewed to evaluate its affect on the integnty of the steam generator tubing. This condition has been found acceptable due to the slight mass of the pin involved. Since it cannot be determined if the rollec: pin was installed by the valve's manufacturer a slight 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 probabilities 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.

NRc Form 366A (6-89) i

-.____-__m__________________I

f .' jj- '^'

t#C Form 366A U.S NUCLEAR REGULATORY COMMIS$10N APPROVEO OMB NO. 3160-01C4

^ (6-80) : EXPIRES: 4/30/92

's~

Estirrcted tharoon por response to comply with this

~. ,

LICENSEE EVENT REPORT (LER)- '",'g,m*4*,a

,, ,, *$,",$4" aug0,f,,h',t ,f,*'*,t'8,,,

n TEXT CONTINUATION .no Reports uan gement erenen (P-63o) U.S. Nuclear Regulatory Commission. Wasnington. DC 20556 and to tne Paperwork Reouction Profect (3160-0104). Office of Manageme'st one Buopet Washington. DC 20503 FACILITY NAME (1) DOCKET NUMf>ER (2) (Em Nuvern ici - PAGE(3i YEAR N Millstone Nuclear Power Station -

Unit 2 - ~

OF 0l 5l Of 0l 0l3 l3 l6 8l7 0l 0l 9 0l1 0l 3 0l3 TEXT in more speco is reawarea. use soaitenai NRc Form 3 sea si tiri 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 outafei no problems of a similar nature were found m these valves. The appropriate valve preventative maintenance procedure has been revised to inspect for the proper installation of this roll pin.

V, Ablitions! Information The Main Feedwater Regulating valve, 2-FW-51B, is a Copes - Vulcan 12" 900* angle style valve assembly with a P-200-12 operator.

Similar LERs: None I

1 l

l  ;

I l

1' l

l i

I l

I t

h NRC Form 366A (6-80)

L- a

_