ML20043F452

LER 90-007-00:on 900517,control Room Received Indication of Half Scram & Isolation of Inboard Reactor Water Cleanup Isolation Valve.Caused by Spurious Trip of Channel a Reactor Protection Sys Epa.Design Change implemented.W/900611 Ltr

ML20043F452
Person / Time
Site:	Hope Creek
Issue date:	06/11/1990
From:	Cowles R, Hagan J Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-007-02, LER-90-7-2, NUDOCS 9006150026
Download: ML20043F452 (5)

Text

't a.

DIMG Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge. New Jersey 08038 Hope Creek Generating Station June 11, 1990

- U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

- HOPE CREEK GENERATING STATION DOCKET NO. 50-354 UNIT NO. 1 LICENSEE EVENT REPORT 90-007-00 This Licensee . Event Report 'is being submitted pursuant to the requirements of 10CFR50.73(a) (2) (iv) .

Sincerely,

.J. H Jan Gener Manager -

Hope Creek Operations RBC/

Attachment SORC Mtg.90-051 C Distribution 9006150026 900611 PDR ADOCK 05000354 S PDC

%p L

- The Energy People 95 2173(%W 12 88 1

WRC t.RM Joe U S NUCLEAR t.E0VL&f 0Av COMMi&$ ION APP 840VID OMS NO 3160 0104

, EXP18't$ 4'30'92 N8 eMATs0N C L (CT ON 8 0 Ett 60 0 H $ FOR AK UCENSEE EVENT REPORT (LER) g u,t,N4sy c,A,ao;= g t g T;M,A,T g Tgt g APE RWOke Rt TION JC (3 0 IC OF MAN AGEMENT AND CUDGET m AlmiNGTON DC 20503 poCati aspesesR 43: eaus i:s, 81CILITV NAast 96 Fort Calhoun Station Unit No. 1 015101010191 A lei 1 IoFln I c; I t#)gt top Potential for Overpressurization of Auxiliary Feedwater Pinino IVINT Daf t 191 Lth humastR sel AGPORT oaf t (?) OTHER f ACILITit$ INVOLWs0 a01 MONTH DAY vtAR vtAR "MU '*, 6 ",8j,$ MONTH DAY vtAR 8Aci6rtv hAwes DOCS.t T NUMethisi N 015l01010 1 1 I 1

~ ~

0] 5 Il 1 90 9l0 0l1l6 0l 0 0 l6 1l1 9 l0 01 5 1 0 to1 0 1 I I

,,g ,,,,,

Twit REPORT is sVSMift:0 PURSUANT TO TMt RkOUIREMENTS OF 10 CFR l Icaer* en* ** *'** of f*e fede**pf Oh

_"* 5 m ==i a *=i.i n ni.H2H .

n eu.i  ;

g _

m .inu nmi.im u ni.aH.i nn= H Hai n.,u.i j

ni 01010 n =NHiHe _

n=NHai _ _ g,m gsg,;aggy,e ,, l 20 4=talf1Het to.73tallittli to 736eH2 Hee 68HAl J86Al l N 40Sisill Havl 90.734a42141 to 731sl(2HwiH51 N eA51 sit 1Het to 73ieH2Hi#si to 73 sit 2Hal LICt8sHE CONT ACT POR TMit LtR 1121 Nivt TELEPMONE NUMetR ARiaC004 Mark Hollinased. Shift Technical Advisnt a In I ? c;l a l l l_ l A i n h 1 1 COMPLtTE ONG LINE FOR 4 ACM CORAPO,'ENT F AILURE DESCRISED IN TMil REPORT H31 MA O t R #

CAult SylT E M COMPONENT C A$'o {,ta ns CAUse $vsttu COMPONtNT "'%$C 0 en k' l I I I I I I I I I f I I I l l 1 I I i l l l I I I I I MONTH OAv vtAR SUPPLEMilNT AgtPORT G RPtCit01941 4tS ll* ree sonwete tKPLCtlO SUOnel3SION CA Til ko l l l ASSTR ACT itsev, M I400 aseres a e , asPremnaemir Areeea s'ap e asete typeweerma maest 06:

Accident scenarios have been identified by which the Auxiliary Feedwater piping from the discharge of turbine driven Auxiliary Feedwater pum) FW-10 can be overpressurized. The scenarios require FW-10 to be running in tie recirculation mode, with no injection to the steam generr.cors, a coincident loss of instrument air, and a single active failure of the pump's speed limiting governor. In accordance with 10 CFR 50.72 tb)(2)(i), this potential for overpressurization was reported to the NRC On May 11, 1990 at 1725 hours0.02 days <br />0.479 hours <br />0.00285 weeks <br />6.563625e-4 months <br /> CDT as a condition outside the design basis.

A design deficency during plant construction which eliminated the specification for an overspeed trip on FW-10 has been determined to be the cause of this condition.

An engineering analysis of tne pump's failure mode has determined that, although the maximum pump discharge pressure will exceed the piping design pressure, catastro) hic failure of the piping or associated components will not occur. Based on tiis analysis, Safety Analysis for Operability 90-012 was approved on May 13, 1990, to justify acceptability of plant operation.

A permanent resolution to the concern for overpressurization of the auxiliary feedwater piping from overspeeding of FW-10 will be completed during the 1991 Refueling and Maintenance Outage.

NRC. - =.iu.,

9 LIONSIE B7NT RElWT FACIIIlY NAME (1) DOGET NUMBER (2) PAGE (3)

IDPE CRHX GENERATING ETATIm 0 5 0 0 0 3 5 4 1 7 4 TITLE (4): INGINEDD SAIElY IB1URIE (ESP) ACIVATIm (RFAC10R WA1TR CLEANUP ISXATIm) IIJE 101 RIPPING OF REACIUR IWITLTIm SYS11N OIANNEL "A" IIICIRICAL IWITLTIm ASSINBLY f% TNT DA1E (5) IIR NLMBER (6) RDWT M1E (7) OnlIR FACILITIES INVOLATD (8)

Unl IAY ** NLMBER ** Umi iTAR YEAR YEAR REV DAY FACILITY NAME(S) DOCKET NUMRER(S) 0 5 1 7 9 0 9 0 -

0 0 7 -

0 0 0 6 1 1 9 0 OPERATING 1 BIIS RDWT IS SUIMITITD IURSUAVI 10 UIE REULHRfMENPS OF 100R: (OIDCK (NE OR P0RE HE10W)(11)

KI10 (9) 20.402(b) , _ 20.405(c) , X_X 50.73(a)(2)(iv) _ _73.71 (b)

KWIR 20.405(a)(1)(1) 50.36(c) (1) __50.73(a)(2) (v) 73.71(c)

._cnER (Specify in IET7L 1 0 0 _

50.36(c) (2) ._

20.405(a)(1)(ii) 20.405(a)(1)(iii) ._ ._50.73(a)(2)(1) ._50.73(t)(2)(vii)Abstract below

\\\\\\\\\\\\\\\\\__ _20.405(a)(1)(iv) ._50.73(a)(2)(ii) ._50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) and in Text) 50.73(a)(2)(iii)

\\\\\\\\\\\\\\\\\ 20.405(a)(1)(v) 50.73(a) (2) (x)

LIONSEI GWTACT FOR 11[IS IIR _(12)

NAME 1T117WNE NUMBER Richant Cowles, Senior Stafi Engineer - Tedmical 60 9 3 3 9 3 4 3 1 (IMPIEIE ONE IJNE 10R FA010:MKNINT FAlll3RE N0 LID IN DlIS RDORT (13)

CAUSE SYSUM EMKNENT MANUFAC- RE2 W TABLE \\\\\\\ CAUSE SYElIM GMfWINT MANUIRC- RDMTABLE 1URIR 10 NPRDS? \\\\\\\ 1URER 10 NPRDS?

B JC IER 0080 Y \\\\\\\

SUPPIIMFNTAL RUMT EXPfURD? (14) YELL I NDlXX DALE 11PErnD (15) Uni DAY YEAR \\\\\\\\\\\\\\\\\\

l l I \\\\\\\\\\\\\\\\\\

ABS 1RAct (16)

On 5/17/90 at 1353, the control room received indication of a half scram and isolation of the inboard Reactor Water Cleanup (RWCU) isolation valve.

The above actions occurred as a result of a loss of the power supply to the Channel "A" Reactor Protection System (RPS) electrical bus when the normal power supply' Electrical Protection Assembly (EPA) experienced a spurious trip. The Channel "A" RPS bus was re-powered from its alternate power source, and the half scram and RWCU isolation were reset. Followup troubleshooting by the Maintenance Department could not determine a definitive reason for the trip of the EPA, however, it is suspected that the trip resulted from an EPA performance problem similar to those noted in General Electric Service Information Letter (SIL) 496. Corrective actions include scheduling of modifications to all RPS EPAs as described in SIL-496.

EPA Manufacturer: General Electric Type: TFJ Part Number: 184C449P001

w  ;

A e

LIGNSEE EVENT RDWT (IDt) 'IIXT ONTDIUATim FACILITY NME (1) DOLET NUMBER (2) 11R NUMBER (6) PME (3) J iTAR ** NLDWER ** REV HOPE CRED: GENERATDG STATEN 05000354 )

90 -

0 0 7 -

0 0 0 2 CF 0 4 1

PLANT AND SYSTEM IDENTIFICATION j General Electric - Boiling Water Reactor (BWR/4)

Reactor Protection System (EIIS Deaignation: JC)

Reactor Water Cleanup System (F11S Designation: CE)

IDENTIFICATION OF OCCURRENCE Engineered Safety Features (ESP) Actuation (Reactor Water Cleanup Isolation) Due to Tripping Of Reactor Protection System  ;

Channel "A" Electrical Protection Assembly Event Date: 5/17/90 Event Time: 1353 This LER was initiated by Incident Report No.90-050 CONDITIONS PRIOR TO OCCURRENCE Plant in OPERATIONAL CONDITION 1 (Power Operation), reactor power 100%, unit load 1110MWe.

DESCRIPTION OF OCCURRENCE On 5/17/90 at 1353, control room personnel received- indication of a half screm and isolation of the inboard Reactor Water ,

Cleanup ( p' jeu) isolation valve (HV-F001).

. The Nuclear Control Operator (NCO, RO licensed) noted that an electrical protection assembly (EPA) for the Channel "A" Reactor Protection System (RPS) normal power supply had tripped. Channel "A" RPS was re-energized from its alternate power source, and the half ^

scram and RWCU isolation were reset. A work request was initiated to troubleshoot the tripped EPA, and- the Senior Nuclear Shift Supervisor (SNSS, SRO licensed) initiated a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> non-emergency report per 10CFR50.72 due to the- RWCU isolation.

APPARENT CAUSE OF OCCURRENCE This occurrence was caused by a spurious trip of a Channel "A" RPS bus EPA.

ANALYSIS OF OCCURRENCE Followup troubleshooting by the Maintenance Department could not determine a cause for the trip of the EPA. Over and under voltage trip setpoints were verified to be within tolerance, and no perturbations from the respective RPS motor generator set were noted.

-x

g i

LICENSEE LVDff RDORT (IER) MT 03frDO. TIN FACILITI NN: (1) 10CKET NUMBDi (2) IIR NUER (6) PME (3)

YIAR \\ NLPEER \\ REN

- HOPE: OtEIK GENERATING STATIN 05000354 0 0 7 -

0 0 0 3 T 0 4 ANALYSIS OF OCCURRENCE. CONT'D t

Systems Engineering reviewed the event, and determined that the EPA trip exhibited characteristics of EPA performance problems similar to those identified in GE SIL-496, which was issued in

-August, 1989.- In response to SIL-496, in February, 1990, Systems Engineering initiated a design change to replace existing logic cards in all EPA's at liope Creek with upgraded logic cards as recommended by GE. This design change is expected to be implemented prior to the end of the stations third refueling outage in early 1991.

PREVIOUS OCCURRENCES ~

There have been 3 previous reportable occurrences initiated by tripping of RPS electrical protection assemblies (reference:

LERs86-007, 87-021, and 89-022). In all cases, the EPAs

, tripped on undervoltage conditions due to either undervoltage L setpoint problems or undervoltage conditions on the alternate power supplies. No " spurious" EPA trips have previously occurred at Hope Creek.

-SAFETY SIGNIFICANCE This incident had ininimal potential safety significance.

Immediately following the EPA trip, RPS channel "A" was ,

re-energized from its alternate power supply.. Technical Specifications permit operation in any operating condition for .,

up to. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with one RPS channel inoperable.

Had RPS channel "B" been inoperable at the time of this occurrence, a s L reactor scram would have occurred, and a reactor scram is bounded by UFSAR analysis.

EQUIPMENT / MANUFACTURER DATA EPA' Manufacturer: General Electric EPA Type: TFJ Part Number: 184C449P001 l

{

\

l l

1

y ..

. LIONSEE IVINF REKRP (IIR) 1RT ONTINATION 17CIIJlY 10ME (1) DOCKLT 10EER (2) IIR NUMIER (6) PAh (3)

YEAR \\ )UEER \\ REN ltFE OHK GENERATDC STATICH 05X)0354 90 -

0 0 7 -

0 0 0 4 OE' O 4 pORRECTIVE ACTIONS The design change as recommended by GE SIL-496 with respect to EPA logic cards will be implemented prior to the end of the stations third refueling outage.

Since ely,

.J 11 Jan Gen Manager -

Hope Creek Operations RBC/

SORC Mtg.90-051 1

1

-- - - _ _ - - _ _ - . - - - _ - _ - _ - _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _