ML19324C104

LER 89-017-00:on 891010,low Level Signals Occurred,Causing Group 1 Isolation,Core Spray Initiation Signal & auto-start of Common Emergency Diesel Generators.Caused by Perturbation of Ref Leg.Procedures revised.W/891106 Ltr

ML19324C104
Person / Time
Site:	Brunswick
Issue date:	11/06/1989
From:	Harness J, Pastva M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BSEP-89-0997, BSEP-89-997, LER-89-017-02, LER-89-17-2, NUDOCS 8911140352
Download: ML19324C104 (8)

Text

V.. . .. .?A..

l .. . .

e-j L

CSS &

) ' Carolina Power & Light Company Brunswick Nuclear Project P. O. Box 10429

.'s Southport, NC 28461-0429 November 6, 1989 p

. FI LE : - B09-13510C 10CFR50.73 i' SERIAL: BSEP/89-0997

,. U.S, Nuclear Regulatory Commission L ATfN: Document Control Desk

-Washington,.DC. 20555

., BRUNSWICK STEAM ELECTRIC PLANT UNIT 2

, DOCKET NO. 50-324 LICENSE NO. DPR-62 LICENSEE EVENT REPORT 2-89-017 Gentlemen:

In accordance with Title 10 to the Code of Federal Regulations,.the enclosed This report fulfills the requirement for

'l Licensee Event Report is submitted.

a written report within thirty (30) days of a reportable occurrence and is in '

accordance with the format set forth in NUREG-1022, September 1983.

Very truly yours, W

J.'L. Harness, G no 1 Manager ,

Brunswick Nuclear Project -

HJP/ meg t

Enclosure cc: Mr. S. D. Ebneter Mr. E. G. Tourigny BSEP NRC Resident Office 4 1

l l

l l

l l

h i

[ h s , ,

~ - }

7 i

o ,

J

• . l OfAC Pere MS W.S. SeuCL4 A A L SOULATORY Cosmasmoose 3,*Rovi0 Due e,0. sienne  ;

'"Na'* "'"

LICENSEE EVENT REPORT (LER) '

1 00cetti sevangen (3) raus as PAJitsTV seamst M Brunswick Steam Electric Plant Unit 2 015 l 0 l 0 l 0 l 31214 1 l0Fl Ol 7 1 TIT'svUnplanned Actuation of Reactor Low Level Instrumentation Due to Suspected Perturbation af Rnctor Instrumentation Lines While Returnine C32-PT-N008 to Service Fn11new Calibration Ov9eet DAf t IS LtR counesen ISI RSPORT oaf t spi OTHER F ACILITits sNvoLvt0 lei DOCILET NvuotRisi MONTH CAV YLAR YEAR MN,'6 ",'f,*,$ MONTH DAt vtAR #acistiv haute 015101010 I I I

~ ~

1l0' 1l 0 8 9 8l9 G l 1l7 O!0 1l 1 0l6 8 l9 0 isto l0 toi 1 i THIS RSPORT IS SUOMITTto PURSUANT TO THE DEQUIREMENTS OF 10 CFR R- (Cneca one e, more er >=e ressew.ael Hu Aflaae opt'.it Me m 5 ,,,,,,,, ,,,,,,, ,, ,,, n, n,,, ,,,,,,,

y .71 0 i 0, 0 n e i.imm m i.imm

.0.w m m

.0.wisim

.0 ni.imm ee nwmi..i n.ru.i gMtgs g . g 30 40btsH1Hual 80.73iaH2H4 to.736eH2Heu4HAl Js4AJ

< J.M _ _

s M 40$leH1Hivi 60.73 te H2 Hul 90.TaneH2Hve44 hoi to destell1Hil 60.73iaH28t 60 to.736eH2H6)

LICEN$lt CONTACT FOR TMIS LER Off gy,yg TELEPHONE NUMetR ARE A CQDS M. J. Pastva Jr., Regulatory Compliance Specialist 91 1 19 415171-12 131115

, COMPLif t ONE LING 90R E Ach COMPONENT F AILURt OtsCRitt0 IN TMit REPORT (13)

C1U58 SY3TtM COMPONENT MQ'g A C. 4i {o0,,R,T,Aj,8 CAvst sysTIM COM*0NENy M A% AC. n{0 N*R I' s s

i I I I I i 1 1 1 I I I I f I l l I I l I l l I I I l l MONTH DAY vtA*

SUPPttMEN.'AL S APORT IRPECTIO f14 SUBMi$$ ION 188 LII yes, ter nosete LKPRCTfD SVOhrIS$10N DA ffl NO l l l Auf a ACT tri,~r to r<w < . te.. . wrsea one.. o.<e evo ~ tr*~ *"* i e'ei At 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br /> on 10/10/89 inadvertent Unit 2 low water Icvel (LL) No. 2 and 3 signals occurred causing a Group 1 isolation, Core Spray (CS) initiation l signal, and auto-starting of the Units 1 and 2 common emergency diesel generators (DGs). The Unit 2 1989-1990 refuel / maintenance outage was ongoing with the Reactor (Rx) defueled. The CS, Residual lleat Removal (RIIR), Rx Water Cleanup (RWCU), Standby Gas Treatment, and the Rx Building ventilation systems l were under equipment clearance. Return to service of Rx feedwater/Rx pressure transmitter (PT) C32-PT-N008 was ongoing. The Control Operator became aware '

of this event through Control Room indication / alarm annunciation. An RilR/ Low Pressure Coolant Injection (LPCI) initiation signal and a Group 3 of the RWCU l System were not confirmed. The isolation / initiation signals were reset and by l 1818 hours0.021 days <br />0.505 hours <br />0.00301 weeks <br />6.91749e-4 months <br /> the DGs were returned to standby. This event had minimal safety I

significance.

1.

l The LL signals are attributed to perturbation of the instrument sensing leg common to N008 and Rx LL instrumentation when N008 was returned to service, por OPIC-PT001. An instruction /precautlonary statement was not provided with the procedure to ensure the test pressure on NOO8 was equal to the system pressure conditions prior to return to service.

l In addition to OPIC-PT001, two similar procedural problems will be resolved by l 12/20/89. Prior to subsequent startup, appropriate testing will assure operability of the Rl!R/LPCI and RWCU Group 3 functions.

go man, va avetims;uctoav ca.wimo=

UCENSEE EVENT CEPORT (LER) TEXT CONTINUATl":N mawso ove no mo-oio.

ixe nie- ewe.

MNV 80 f4 oOCR41 NVMeth til LIR NUM88R lel P&OS 84)

Brunswick Steam Electric Plant

,,.- am m Unit 2 o ls lo l0 l0 l 3l2 l 4 8 l9 -

0l 117 --

0l0 0l2 0F 0l7 w . a e,samenwanc==mamm Event Inadvertent Reactor low water level No. 2 and 3 signals caused a Unit 2 Primary Containment (PC) Group 1 isolation of Reactor water sample PC outboard isolation valve 2-B32-F020 Reactor Core Spray initiation sf3nal, and automatic starting of the Units 1 and 2 common emergency alternating current diesel generators Nos. 1-4. The low level signal is attributed to unplanned actuatica of Reactor level instrumentation due to a suspected perturbation of Reactor instrumentation lines during the return of Reactor feedwater/ Reactor pressure instrument C32-PT-N008 to service following routine instrumentation calibration.

Initial Conditions i The Unit 2 1989-1990 refueling / maintenance outage was in progress with the Reactor (EIIS/AC) defueled and the Reactor cavity (EIIS/DF/**) flooded. The Reactor Core Spray Sy' stem (EIIS/BM), Residual llent Removal (RHR) (Ell) System (EIIS/BO), Reactor Water Cleanup (RWCU) System (G31) (EIIS/CE) Standby Gas Treatment System (EIIS/BH), and the Reactor Building Ventilation System (E1IS/VA) were under equipment clearance for ongoing outage activities. In addition, the Units 1 and 2 common emergency alternating current (AC) diesel generators (DGs) (EIIS/EK/DG) acre in standby readiness and normal off-site  ;

power was available. Maintenance technicians were in the process of returning Reactor feedwater (C32) (EIIS/SJ)/ Reactor pressure transmitter (PT),  !

2-C32-PT-N008 (EliS/SJ/PT) to service following a periodic preventative maintenance calibration of the instrument in accordance with the process i instrument calibration (PIC) of General Electric type 551 and 552 pts, OPIC-PT001.  !

• • E11S component description unavalloole Event Description  !

At 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br /> on October 10, 1989, when the technicians opened the instrument ,

isolation valvo, 2-C32-PT-N008-3 (EIIS/SJ/ISV) in accardance with Step 7.4.3 of- l the procedure, Reactor low water level Nos. 2 and 3 signals occurred. These low level signals initiated a Primary Containment Group 1 isolation signal, a l Core Spray initiation signal, and automatic starting of the units' emergency l

I AC DGs. Per design, the DGs did not tie onto their respective emergency (E) 1.

buses, El-E4 (EIIS/EK/BU) as normal off-site power remained available. The 1 Unit 2 Control Operator (CO) became aware of this event through appropriate l Control Room indication and alarm annunciation. Following this event it was verified that the Reactor water sample outboard primary containment isolation valve (PCIV), 2-B32-F020 (Ells /AC/ISV), which had been opened prior to the l

event, had closed as required due to the Group 1 signal. At 1533 hours0.0177 days <br />0.426 hours <br />0.00253 weeks <br />5.833065e-4 months <br />, following verification that the Reactor low level alarm annunciations were raise, the isolation / initiation signals were reset. In addition, the DGs were

- a m m. .9eoeoieseo.eu useas

,D '.

u ,

g' mes , va =veta= c novutoav co .o.,

UCENSEE EVENT REPORT (1.ERI TEXT C3NTINUATl2N emevo ose no mo-oio.

no. ass: emra pausserv geAug gu Dockte NUMDea W gga asynsega iel paos tai Brunswick Steam Electric. Plant *taa -

W.L 'L*a Unit 2 o l6 l 0 l 0 l 0 l 3l 2 l4 8l9 -

0 l 1l 7 -

010 0 13 0F 01 7 gr' asum aim 4. ==*=t es. .mmmw mic a== ama w im t

secured from operation and returned to standby readiness (DGs Nos. I and 2 at 1610 ho2rs) (DGs Nos. 3 and 4 at 1818 hours0.021 days <br />0.505 hours <br />0.00301 weeks <br />6.91749e-4 months <br />). It could not be determined if the Wir4/LPCI signal had been received at the time of the event. In addition, as the WCU Sys6cm was unae q"ipment clearance with the System inlet outboard PCIV, 2-G31-F004 (EIIS/CE/ISV) closed, visual confirmation of the required closure signal to the valve, due to the incurred low level No. 2, was unavailable.

Event Investigation Following the event, interviews with the involved maintenance technicians were conducted in order to reconstruct their actions relative to the event. Based upon those interviews, it was concluded that the technicians had properly adhered to the procedure. Therefore, personnel error by the involved technicians was ruled out as a root cause or contributing factor to the event.

The procedure isolates the instrument from its respective instrument sensing line (EIIS/SJ/PSX) through closure of the instrument isolation valve, 2-C32-PT-N008-3 (EIIS/SJ/ISV), in order to perform a five point (0*., 25%, 50%,

75*., and 200'i,) calibration of the instrument at corresponding ' applied test pressures ranging from 814-1114 pounds per square inch gauge (psig). At the time of the event, with only the preexisting static hydraulic head pressure on the instrument sensing line, the instrument had been satisfactorily calibrated following application of the 1114 psig test pressure and the technicians were in the process of reopening 2-C32-PT N008-3 in order to return the instrument to service in accordance with Step 7.4.3 of the procedure.

An adequacy review of OPIC-PT001 determined the procedure did not provide for the contingency of valving in the transmitter to a depressurized reference icg. Consequently, when 2-C32-PT-N008 was valved back into service the calibration test pressure was applied to the already depressurized instrument sensing leg.

The investigation concluded the low level signals resulted from actuation of Reactor low level instrumentation on the common referencing Icg with 2-C32-PT-N008, due to a momentary perturbation of the leg when the instrument was returned to service. Instrumentation, which shares this common reference sensing Icg, is listed in Table 1.

irac roar see. .uaopo teseo.e24assass

.r __

g' ewe ssen u a. Nuctimsoutt.fomy commission l

c' UCENSEE EVENT REPORT (LER) TEXT CONTINUATGN <*novio ove =o. mo.oio4 )

sxPines; awe F451b4TT014A8 On Docat? NVM8th us LER NUMB 4R tes Paot tal B'runswick Steam Electric Plant 'f'" " M a" M*siE Unit 2 menans.

0 l5 l0 l0 l0 l 3l2 l 4 8l 9 -

0 l1l7 -

0l0 0l4 0F 0l7 emenw w.amannmacwassawnn I

i

-During a review of plant documentation to c.samine the applicability of prior j similar occurrences to this event, the appropriateness of corrective actions for LER l-87-017 as they relate to this. event were evaluated. The event ,

E reported in LER i-87-017, which occurred on June 12, 1987, involved valving-in }

depressurized test equipment to the p.esaurized instrument reference leg common j to 2-C32-PT-N008. The investigation of the event in 1987 determined it had ,

resulted from a momentary perturbation of the reference leg which caused an i automatic Reactor Scram due to. actuation of the affected Reactor level instru-montation. As part of corrective action to the event, a review / revision was conducted of plant procedures involving valving-in instruments on a common reference sensing leg. This effort was focused to ensure the 4.nstruments are properly pressurized prior to valving-in during Reactor power operation.-

- Reexamination of the approach taken in achieving the corrective actions of LER 1-87-017 has determined the actions did not address restoration to a depres-surized reference leg, in that only procedures for valving-in devices to a pressurized Reactor S'ystem were considered.

The investigation determined that the most probable rocson why an RHR/LPCI initiation signal was not received as a result of this event is the suspected relatively short duration of the Reactor low level signal, which did not allow both the RHR Reactor low level channel relays, 2-Ell-K7B and K8B (EIIS/B0/RLY), and the RHR/LPCI seal-in relay, 2-Ell-K9B (EIIS/B0/RLY), to energize for a sufficient time to allow the normally open K6B seal-in contacts to close.

Action Taken The failure to review procedures in order to provide for return to service condition regardleas of Reactor pressure, should be corrected through improved I root cause/ corrective action determinations at the Brunswick station. The

! corrective action program is viewed as an appropriate measi.re to ensure that adequate attention is devoted to ensure the proper identification and focus of corrective action.

l.

l As a result of this event, a review was conducted to identify Reactor instru-l mentation and maintenance procedures involving instrument sensing lines which l are sensitive to pressure porturbations, in order to ensure adequate procedural controls are provided regarding a pressurized or depressurized Reactor System.

I

This effort revealed two additional instrumentation calibration instructions were inadequate to ensure the Reactor System pressure is taken into account l

l during return of the involved instruments to service.

l l

l l

l l NRC PORM agen e U B opo iO44.o.624 634 486 C, __ _. .

)

l5100 persa m U S. NVCLla% Kt0ULATORY COMMise#71 UCENSEE EVENT REPORT (LER) TEXT CONTINUATION u.aovt0 0Me =o 3 iso..oio. -

EXPIAES: t/31C

'~

FABIEITV 80AM I4 00CF.87 NVMSER 83) LER NUMBtR (el PAtt 130 Brunswick Steam Electric Plant "'" " Ea ' dan Unit 2 o l5 l0 l 0 l 0 l 31214 819 -

0l117 -

0l0 0l5 0F 0 17 1 mmaan ass,emesses .ammemac=menwon

)

l l

i Actions to be Taken By December 20, 1989, OPIC-PT001, as well as the two additional procedures identified during the subject procedural review, will be appropriately revised to provide for consideration of the state of the Reactor System (depressurized or pressurized) during valving-in of instrumentation or other test equipment /

devices. ,

In additien, prior to startup of Unit 2 from the current refueling /msintenance outage, appropriate testing will be performed to ensure operability of the RHR/LPCI actuation instrumentation. It is felt the Group 3 isolation capability of the RWCU System will be verified through postmaintenance testing

! requirements which follow the cancellation of the involved equipment clearance on the system.

Event Assessment The occurrence of this event under other credible alternative conditions would not have been more severe.

In addition to LER 1-87-017, prior events involving instrument perturbations while valving-in instrumentation have been reported in LERs 1-85-039, 1-35-047, and 2-86-020.

l I

I I

l i

l l

l l

l l'

l l

l

\

l i

l l

l 1'

l NAC # Em seen e U S OP01946 0 624 634 466

m Sene W U 5. NUCit!& EE;UL170av COMMISSION  !

UCENSEE EVENT CEPORT (LER) TEXT CSNTINUAT!ON **aovio ove =o. mo-oio.

EXPentS. S/31/03 paggEpry eneass fll 00CMlfNVMD4R W (ggg,yggggqq) p (g (33 Brunswick Steam Electric Plant 'f*" "UA' a - 7 02

', Unit 2 0 l5 l0 l0 l0 l 3l2 l 4 819 -

0l1l 7 -

0 l0 0 l6 0F 0 17 nur asse anse e asses ass sesamw asse amin new im

'n TABLE 1 Instrument No. Function B21-LT-N025A-1 Initiates signal for closure of main steam

'(EIIS/SB/LT) isolation valves (MSIVs) (EIIS/SB/ISV),

2-B32-F020, main steam line drain outboard PCIV  !

2-B21-F019 (F.IIS/SB/ISV), RWCU System inlet l vutboard PCIV, G31-F004 (EIIS/CE/ISV), isolation l' ,

of the Reactor Building Ventilation System, and ,

automatic starting of the SBGT trains A and B j (EIIS/BII/**).  ;

B21-LT-N025B-1 '

Initiates signal for closure of MSIVs, (EIIS/SB/LT) 2-B32-F020, 2-B21-F019, G31-F004, isolation of  ;

the Reactor Building Ventilation System, and automatic starting of the SBGT Trains A and B.  ;

i B21-LT-N025A-2 Initiates an anticipated transient without a

'(EIIS/SB/LT) scram (A'IVS) (EIIS/JC) trip of. reactor recirculation pump A (EIIS/AD/P).

B21-LT-N025B-2 Initiates an ATWS trip of reactor recirculation (EIIS/SB/LT) pump B (EIIS/AD/P).

B21-LT-N031B Initiates automatic initiation of the Reactor (EIIS/SB/LT) Core Isolation Cooling (RCIC) System (EIIS/BN), l permissive for actuation of the Automatic Depressurization (ADS) System (EIIS/*),

initiation of the RHR low pressure coolant injection (LPCI) mode, Reactor Core Spray System, and automatic starting of units' common emergency AC DGs.

B21-LT-N031D Initiates RCIC System automatic initiation, (EIIS/SB/LT) permissive for actuation of the ADS, initiation of the RilR/LPCI mode, Reactor Core Spray System, and automatic starting of the units' common emergency AC DGs.

B21-PT-N021B *** Provides 410 psig Core Spray injection (EIIS/SB/PT) permissive as well as permissive for the 310 psig Reactor Recirculation pump discharge valve closure.

l l

l l

I

,,ne . . .U a o o.iese.o.e24 sw4ss

. . **88 -- . _. .-. _ ._. ,__ --_

v s avetim mumo , co==.mio.,

deseen

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION u*aoveo o o. w-oio4 i , sx,,au. sous

. - N, ooe m v .. m t,,,,,,,,,,,,, ,.o. .. 1

.....~,.., .. ....

1 Brunswick Steam Electric Plant '

Unit 2

- querees eseeosmew ensumwesemannawim o l5 l 0 l o l o [3 l 2 l4 8l9 -

0 l1l 7 -

0l0 0l7 or 0l7 !

l TABLE 1 (Cont'd)

B21-IT-N0230 *** Initiates the Reactor Vessel liigh pressure  !

(EIIS/SB/PT) automatic scram B21-PT-N023D *** Initiate;/ the Reactor Vessel liigh pressure

.(EIIS/SB/PT) automatic scram

• EIIS system description not available .

• 5'EIIS component description not available

• • • It is believed the magnitude of the incurred sensing leg perturbation was

'not sufficient to have affected this instrument. ,

?

.. .v i ono.ieu.o.e:4 ea4se

, . . - . . ~ . _ . . . , . , . . - - - . . . - . . . - . - , , . . . . , . - . . _ . . , .- . - - -. - - . - - ..,,. -