ML20006E116

LER 90-001-00:on 900118,during Functional Surveillance on Steam Generator Pressure Channel 526,channel 525 Spiked Low, Causing Reactor Trip & Safety Injection.Caused by Failure of Pressure Transmitter.Transmitter replaced.W/900126 Ltr

ML20006E116
Person / Time
Site:	Byron
Issue date:	02/09/1990
From:	Gierich T, Pleniewicz R COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BYRON-90-0108, LER-90-001-03, NUDOCS 9002220074
Download: ML20006E116 (4)

Text

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/ ) B ron Nucint Station

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Nv,/ - Byron, Illinois 61010 -

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-January 26, 1990 l> -Ltra . BYRON 90-0108.

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

l The enclosed Licensee Event Report from Byron Generating Station is being E transmitted to you in accordance with the requirements of 10CFR50.73(a)(2)(lv).

i This report is number 90-001; Docket No. 50-455.  ;

a Sincerely,

-i Yi  !

.[V _ R. L Plenlewicz -

Station Manager Byron Nuclear Power Station l 1

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Enclosure Licensee Event Report No.90-001

cc A. Bert Davis, NRC Region III Administrator -I W' Kropp, NRC Senior Resident Inspector l INPO Record Center

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i 9002220074 900209  ;

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6 ADOCK 05000455 PDC ,

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LICENSEE EVENT REPORT (LER)

Form Res 2.0 (ccll'ity Name (1)

Docket Number (2) _fage (3)

Byron. Unit 2 01 51 01 01 01 41 51 5 1lof!0!_1_

Title (4)

_RLAC10R TRIP QHJO STEAMLINE PRESSURE DUE TO ONE CHARNEL SPIKIMG LOW DURING A FUNC110N.6L SURVEILLA

_EysttLDate ($1 LER J4 umber (6) _ _RepnLDate (7) Other Facilities Involved (8)

Month Day Year Year fj/j// Sequential ,/j// Revision Month Day Year Facility Names Docket Numberft)

/// Number ///, Number

~

NONE 0151010101 l l

~~~

QJ 1 1 18 910 91 0 0 1 0 11 010 012 Ol 9 91 0 0151010101 ( l' OPERATING (Check one or more of the followino) (11) 1 20.402(b) ___ 20.405(c) l_ 50.73(a)(2)(tv) ___ 73.71(b)-

POWER 20.405(a)(1)(1) 50.36(c)(1)

__ _ _._ 50.73(a)(2)(v) _ 73.71(c)

LEVEL 20.405(a)(1)(ll) 50.36(c)(2) 50.73(a)(2)(vii)

(10) l9l9 _

20.405(a)(1)(tii) _

50.73(a)(2)(i)

_ 50.73(a)(2)(viii)(A)

___ Other (Specify (a Abstract

/////,///////,/,/,//,/////,/////, _., 20.405(a)(1)(iv) ._._ 50.73(a)(2)(ii) ___ 50.73(a)(2)(viii)(B) below and in

////j///////j'/j' j' /j' /////'j /////'j

// / _ 20.405(a)(1)(v) _ 50.73(a)(2)(iii) _ 50.73(a)(2)(x) Text)

LICENSEE CONTACT FOR THIS LER (12)

Name TELEPHONE NUPSER AREA CODE

. T. Gierich. Doeratino Enoineer Ext. 2218 8l115 213141 l514141 COMPLETE ONE LINE FOR EACH COM ON N FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER TO NPRDS TURER TO NPRDS X SlB l l Pl T I l 21 01 4 Y I I I i 1 1 I l l l l Jl I, I I l i 1 I l-SUPPLEMENTAL REPORT EXPECTED (14) Expected Month l pg M ear Submission lyes (If_ythmjete EXPECTED SUBMISS 1011 DATE) #

X l NO 1 l ll ABSTRACT (Limit to 1400 spaces, i.e. approximately fif teen single-space typewritten lines) (16)

On January-18, 1990. Unit 2 was operating at 99% power. An Instrument Maintenance Technician was performing a functional surveillance on steam pressure channel 526 when channel 525 spiked low. A reactor trip and safety injsction followed.

Th2 most probable cause of the event was the failure of the 525 pressure transmitter. The transmitter was rep 1:ced, and the spiking has not recurred. This is the first time the coincidence was satisfied during a pr3ssure transmitter f ailure and resulted in a reactor trip.

This event is reportable per 10CFR50.73(a)(2)(iv) for any event or condition that resulted in manual or automatic actuttion of any Engineered Safety Feature, including the Reactor Protection System.

(0516R/0059R)

r' :

.p-LICENSEE EVENT REPORT (LER) TEXT CQMUHUATION' Form Rev 2.0  :

. FACILITY NAME (1) DOCKET NUMBER (2) JJfLNUMBER (6) Page (3)

Year /// Sequential Revision

,, //

, fff

/// . Number

/j//

ff

/ Number Jyron. Unit 1 0 l 5 1 0 1.jLj .Q_jJllj_1 9 1 0 -

01.0l1 - 010 012 0F Ol 3

TEX 1 Energy Industry Identification System (EIIS) codes are identified in the text as (XX)-

A. 11 ANT CONDITIONS PRIOR TO EVENT:

~ Event Date/ Time 1/18/90 / 0042 ,

Unit 2 H0DE 1 - Power Operations _ Rx Power _o9% RCS ( AB) Temperature / Pressure _HQDnA]_0Reratino

~

B. bESCRIPTIONOFEVENT:

At-the start of this event, steam pressure channel 526 (corresponding to the 2B steam generator) was in test for calibration under 2 BIS 3.2.1-015, Surveillance Functional Test for Loop 526 Steam Generator 2B Pressure Protection Channel III. Limiting Condition for Operation Action Requirements (LCOAR)

.280$ 3.2-la and 200$ 3.3.6-la we re in ef fect. The surveillance placed the channel's protection blstables i in the tripped condition. At 0042 on January 10, 1990, steam pressure channel 525 (corresponding to the  ;

2D steam generator) spiked low creating a 2 out of 3 coincidence in 1 out of 4 main steam lines for a low

~ steam line pressure safety injection / reactor trip. Procedure 2BEP-0, Reactor Trip or Safety Injection (SI) was entered. At step 32 in 2BEP 0, the safety injection termination criteria was met and 2BEP ES 1.1 SI Termination procedure was entered. The safety injection was reset at 0050. All safety systems functioned as designed.- All operator actions were correct.

This event is reportable under 10CFR50.73 (a)(2)(iv).

C. CAUSE OF EVENT:

1he proximate cause of the event was a low spike on a transmitter combined with a coincident channel in test. The root cause cf the transmitter spike is indeterminate.

The Inst'rument Maintenance Technician perfonning the surveillance on channel 526 was several feet away from the protection cabinet (2PA03J) signing a step in the survel11ance package when the trip occurred.

The surveillance was_ essentially complete except for restoring the blstables. Thetehnician'sactivities did not affect channel 525 spiking.

D .~ 1AFETY ANALYSL$:

There was no effect on the health and safety of the plant or public. The plant responded nonnally following the trip and all Engineered Safety Feature Systems actuated properly. The steam pressure loops supply logic to the-water hammer prevention system and initiate safety injection on low steam pressure and steamline isolation on high negative steam pressure rate. Two of the three loops must respond for any

safety actuation to occur.

E.- [DRRECTIVE ACTION 1:

tNR B73382 was written to investigate the spike on the 525 steam pressure channel (2PT-0525).

During troubleshooting, a variety of checks were performed on the equipment associated with this loop.

The transmitter input loop power supply, both isolated and non-isolated outputs, power to the transmitter, the signal comparator, and both lead lag amplifiers were checked and found functioning properly.

.The pressure transmitter was replaced as a conservative measure under this NWR, PNR B73400 was written to monitor transmitter 2PT-0525. No abnormalities were noted. Based on the. lack of further spikes, .the transmitter replacement is considered to be adequate corrective action.

[(0516R/0059R)

o LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rev 2.0 eFACI[ITYkA'ME(1) DOCKET NUMBER (2) LER NUMBER (6) Pace (3)

. Year- /

j/((

Sequential /j{

/

/j/J Revision

// Number Number

.BJtan. Unit 1 0 l 5 l 0 1 0 l 0 l 41 5.1 4 910 -

0_l 0 l 1 - _0__j 0 01 3 0F 01 3 TEXT Energy Industry Identification System (E!!S) codes are identified in the text as (XX)

F. EREVIDU$ OCCURRENCE 1:

No previous LERs were attributed to pressure transmitter failures. A Nuclear Plant Reliability Data System (NPRDS) Component Failure Analysis Report (CFAR) showed a slightly higher f ailure rate (f allures / component-hour) for Unit 1 (2.38E-05) as compared to the industry (7.24E-06). Unit 2 showed a similar comparison (which excluded Unit I f ailures) of 1.34E-05 failures / component-hour compared to an industry rate of 7.65E-06. The significance criterion for the Unit I comparison was 2.642 (which is slightly higher than the 1.645 standard used to indicate higher than industry average) and 0.949 for Unit 2. A review of the NPRDS search did not reveal a common mode failure mechansim. A Total Job Management (TJM) search of the Station's maintenance history identified 35 equipment identification numbers with ITT Barton #763 pressure transmitters. A total of 65 work requests were found. Work request 73382 was the only work request found on 2PT-0525. No common mode failure was identified from the TJM search.

G.  !

(QMPONENT FAILURE DATA:

HANUFACTURER NOMENCLATURE MODEL NUMBER ITT Barton Pressure Transmitter 763 e

1 1

1 l

l l

(0516R/0059R)