ML20012B552

LER 90-001-00:on 900206,full Reactor Scram Occurred During Instrument Surveillance Testing.Caused by Actuation of APRM E Trip Circuitry.Shutdown Margin Revised & Caution Card Placed on Control Room Bench board.W/900308 Ltr

ML20012B552
Person / Time
Site:	LaSalle
Issue date:	03/08/1990
From:	Diederich G, Dorwick K COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-001-08, NUDOCS 9003150291
Download: ML20012B552 (6)

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Common:rith Edison

' LaSalle County Nucl(ar Station

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-- Rural Route #1. Dox 220 .

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March 8, 1990 Director of Nuclear Reactor Regulation U.S.' Nuclear' Regulatory Commission Mail Station Pl-137 Washington, D.C. 20555'

Dear Sir:

i Licensee Event Report #90-001-00, Docket #050-374 is being submitted to your office in accordance with 10CFR50.73(a)(2)(iv).

4 J d ic tation Manager LaSalle County Station  ;

GJD/KCD/kg Enclosure xc: . Nuclear Licensing Administrator NRC Resident Inspector j NRC Region III Administrator INPO - Records Center l

l Q315A g 90030g y l S K 05000374 ( ( l PNU. \

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1 LICENSEE EVENT REPORT (LER) p 7, j Docket Neber (2) Pace (3) l Facility N me (1)

L W 11e County Stat' ion Unit 2 015101010131714 1lof!0l5 Title (4) .

Reactor Scrm durine Instroent Surveillance Testino Caused by Spurious Spike on Averace Power Ranoe Monitor 1 l

Event Date (5) LER Neber (6) Report Date (7) Other Facilities involved (8)

Sequential Revision Month Day Year Facility N aes Docket Weber (s)

Month Day Year Year ///

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/// Haber /// Neber j 01 51 01 01 01 I I ,

01 2 01 6 91 0 91 0 01011

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01 0 013 01 8 91 0 01 51 01 01 01 l l j I

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR

"^ (Check one or more of the followino) (11) i M00E N 1 20.402(b) _ 20.405(c) _X_ 50.73(a)(2)(lv) 73.71(b) l 20.405(a)(1)(l) 50.36(c)(1) 50.73(a)(2)(v) ,,,,, 73.71(c) j POWER ,,, _

20.405(a)(1)(11) 50.36(c)(2) 50.73(a)(2)(vil) other (Specify j LEVEL g  ; ,_ _ ,_,_ ,,

Ii 01 0 20.405(a)(1)(111) 50.73(a)(2)(l) 50.73(a)(2)(vill)(A) in Abstract (10) ,,__ ,,,,_

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LICENSEE CONTACT FOR THl5 LER (12)

I TELEPHONE NUSER Nme AREA CODE Kevin C. Oorwick. Technical Staff Enoineer, extension 2705 8 l 1 15 315171-l6171611 )

COW LETE ONE LINE FOR EACH CO WONENT FAILURE DESCRIBE 0 IN THIS REPORT (13) {

CAUSE SYSTEM COMPONENT MANUFAC- REPORTA8LE CAUSE SYSTEM COMPONENT MANUFAC. REPORTABLE TO NPRDS TURER TO NPROS TURER X tI o l l I I I I N I I I I I I I X Al A l l l Gl01810 Y l l l l l l l Expected Month l Day l Year SUPPLEENTAL REPORT EXPECTED (14)

Subalssion

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Dak ( W 016l210!910 YlYes (if ves complete EXPECTED SUBMISSION DATE) l NO A8STRACT (Limit to 1400 spaces, i.e, approximately fifteen single-space typewritten lines) (16)

On February 6,1990 at 0926 hours0.0107 days <br />0.257 hours <br />0.00153 weeks <br />3.52343e-4 months <br />, while Unit 2 was in Operational Condition 1 (Run) at 99.85 power, during the perforinance of LaSalle Instrument surveillance LIS-NR-403, " Unit 2 Average Power Range Monitor (APRM) Rod j Block and Scr m Functional lest," a full reactor scrm occurred. Morinally the surveillance only causes half-scr ms. At the time of the occurrence, F APRM was trippsd, per the procedure, which tripped Reactor Protection System (RPS) Channel A. While the RPS Channel "A" half scr m condition was in effect, E APRM spiked spuriously, causing RPS Channel B to trip and a full reactor scrm occurred.

Additionally, it was also deterinined that all other expected automatic actions occurred as expected including Primary containment isolation signals when reactor water level reached 12.5 inches decreasing.

Initiation of the event was not due to an actual transient on a parameter which is monitored to protect the reactor core but due to spurious spike of APRM E. Troubleshooting E APRM will continue under Work Request L%857 in an attenyt to deterinine the cause of the spurious spikes.

Unit I was not affected by this event.

This event is reportable pursuant to 10CFR50.73(a)(2)(lv) due to an actuation of an Engineered Safety Feature.

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• J LICENSEE EVENT REPORT (LER) TEXT CONTINUATION' Fo m Rev 2.0 Paes (3) $

. FACILITY NAfE (1) DOCKET NU SER (2) LER NUSER (6)

Year /

j/f Sequential //j/, Revision

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% 11e Coucty Station unit.2~ 0 1 5 1 0 1 0 1 0 1 31 71 4 9I0 - 0 1 0 l'1 - 0I0 '01 2 0F 01 5 l Energy industry identification System (Ells) codes are identified in the text as [XX] l TEXT l' PLANT AND SYSTEM IDENTIFICATION General Electric - bolling Water Reactor 3 v ,

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.EnergyIndustryidentificationSystem(Ells)codesareidentlfledinthetextas[XX].

A.- CONDITION PRIOR TO EVENT Event Time: 0926 Hours Unit (s): 2 Event Date: 02/06/90 Reactor. Mode (s): 1 Mode (s) N ee: Run- Power Level (s): 99.85

, B. DESCRIPTION OF EVENT On February 6, 1990 at 0926 hours0.0107 days <br />0.257 hours <br />0.00153 weeks <br />3.52343e-4 months <br />, while Unit 2 was in operational condition:1 (Run) at 99.85 power,.

• during the perfomance of LaSalle Instraent Surveillance L15-NR-403, " Unit 2 Average Power Range n ~ Monitor. (APRM) Rod Block and Scrm Functional Test," a full reactor scram occurred. Nomally the surveillance only causes half-scrams. Atthetimeoftheoccurrence,FAPRM[lG)was tripped, per the

. procedure, which tripped Reactor Protection System (RPS, C71) [JC] Channel A._ While the RPS Channel "A" ,

half scrm condition was in effect, E APRM spiked spuriously, causing RPS channel B to trip and a full reactor scr m occurred.'

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- After the automatic scrm, the control Room Operator.(NS0) noticed that Control Rod Orive (CRO) (RD, ~ '

Cll) [AA] 26-47 was-latched at the."02" position. .The rod was subsequently manually inserted to the required "00" position. Subsequent testing revealed that control rod 26-47 initially,went to the full-in position. Upon resetting the automatic scr m ,' the rod drifted and latched at the "02" position.

Additionally, it was detemined that all other expected automatic actions operated correctly _ Including' .

Primary Containment Isolation signals (PCIS, PC) [JM] when reactor water level reached 12.5 inches '

decreasing.

Unit I was not affected by this event.

This event is reportable pursuant to 10CFR50.73(a)(2)(lv) due to an actuation of an Engineered Safety.

. Feature.

, C. APPARENT CAUSE OF EVENT -

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The cause of the reactor scram was due to an unexpected actuation of the E APRM trip circuitry. A half scram condition existed due to the perfomance of LIS-NR-403. When the E APRM trip circuitry actuated, a fu11 reactor scram occurred. The cause of the spike is unknown.

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!~ LICENSEE EVENT REPORT (LER) TEXT CONilWUATIDW Fom Rev 2.0 DDCKET NUSER (2) LER WUSER (6) Pace (3)

FACILITY NAE (1)

Year //,/ Sequential // Revision j/,/j/ Number

/j/j/j/ thauber t e lle Countv Station unit 2 0 1 5 1 0 1 0 1 0 1 31 11 4 9IO - 0I011 - 01 0 01 3 0F 01 5 TEXT Energy Industry identification System (Ells) codes are identified in the text as (XX]

.C. APPARENT CAUSE OF EVENT (Continued)

Special recorders had been installed in the RPS logic strings to attanipt to help pinpoint the source and collect data on the character of very short trip signals entered into the RPS logic (LER 314/99-011-41). The recorder on the A2 logic string recorded this scram event. This recorder trace shows that the RPS logic string was opened for slightly less than 1 AC cycie, or approximately 12 - 15 milliseconds. The RPS contactors completely de-energized as expected.. A recorder trace was also obtained from the 81 logic string recorder. It was triggered at the time of the intentional 1/2 scram-from APM F. This recorder trace does not extend in time to the APM E trip. (Due to the scan rate, a recorder trace only shows about 400 milliseconds of data - the E APM trip was 1.3 seconds after APM -j F). However, it does show that the electrical conditions in the B RPS channel were completely stable within 2 AC cycles after the F APM trip, well before the APM E trip occurred.

The RPS recorders indicate that the APRM E trip was internal to the APM drawer, and not induced by the actuation of other RPS circuitry. There appears to be no direct relationship between the survelliance in APRM F and the APRM E trip. This was verified by the testing sequence following the scram, where the functional tests for APM F,'then E, then F again were conducted, with no trip indications appearing in APM E. (Thistestsequenceisthesameasoccurredjustbeforethereactortrip.)

Troubleshooting CRD 26-47 indicated that it is runnlag " hotter" than most other CRD's. This is consistent with data collected over the past year. This data, collected previously per LaSalle Special Test, LST-89-IS1, "CRD Thenuocouple Direct Measurement at 1(2)Cll-R018 Recorder," shows CR0 26-47 running consistently at 350 - 400 degrees F. This high tanqperature condition is likely due to foreign material blockage of the CRD cooling water orifice inside the drive mechanism., This postulation is supported by the fact that the insert stall flows for this CR0 have steadily decreased over the current operating cycle indicating a restricted cooling water orifice. The CRD consistently running " hot" with a lack of coo) lng water flow is believed to have led to greatly accelerated wearing of the stop piston seals, potentially aggravated by nonnal weekly exercising of the CRD and a " hot" scram which occurred in August, 1989. This stop piston seal wearing is demonstrated by the steadily and sharply increasing withdraw stall flows observed over this operating cycle. At the start of this cycle, both the insert and withdraw stall flows were still in the nonnal range. This CR0 was installed in Unit 2 during its first refuel outage 'in the spring of 1987.

Due to the station's awareness of this CRD's high temperature and degrading stall flow trends several corrective measures were attempted during the August,1989 Unit 2 shutdown. This included flushing the CRD in accordance with LaSalle Operating Procedure, LOP-RD-19, "CRD Flushing," which involves a high pressure extended withdraw stall flow to flush collet area. Also, the CR0 cooling water orifice was back flushed through the insert line vent valve in an attempt to increase cooling water flow. Welther of the above actions proved successful for this CRD.

D. SAFETY ANALYSIS DF EVENT Initiation of the event was not due to an actual transient on a parameter which is monitored by RPS.

With the exception of Control Rod Drive 26-47, all systems required to operate functioned as designed.

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LheS""1 EVENT REPORT (LER) TEXT CONTINUATION Fom Rev 2.0 l DOCKET NupWER (2) LER tRpWER (6) Pee (3)

FACILITY NAM (1)- '

' Year /// . Sequential Revision N/ //g// Netper f

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0 l'0 l 1 0lo 0 i s 1 0 1 0 1 0 1 31~71 4 910 01 4 0F 01 5 laSalleCount'yStationunit2 - -

iEXT . Energy Industry identification System (Ells) codes are identified in the text as [XX]

D.- SAFETY ANALYSIS OF. EVENT (Continued)

A shutdown margin calculation was performed by General Electric to detennine the consequences of CRD 47 settling at position "02" lestead of "00." The results indicate that the shutdown margin was 1

. adequate with this rod at position "02" and the next strongest rod fully withdrawn.

-Based on the above, the safety consequences of this event are considered minimal. .

-E.. CORRECTIVE ACTIONS An investigation was performed following the event and the following items were noted:

No radio transmissions were noticed just prior to or during the scram. This addressed any radio frequency noise concerns which may have caused interference with the RPS or APRM drawer's trip ,

logic.

No personnel were present inside the Control Room panels which could have inadvertently bumped a cable.-

An interview of all personnel working in the Control Room area was perfomed and it was detemined '

that no work was in progress that could have resulted in a trip of APRM Channel E. It was also detemined that no one was present in the Cable Spreading Room which could have used a radio.or bumped a cable, it was detemined that no welding was in progress at the time of the event that could have resulted in the trip of APRM E.

Troubleshooting E APRM will continue under Work Request L96857 in an attempt to detennine the cause of the spurious spikes. A supplement to this report will include the results of this troubleshooting and will be tracked by Action item Record (AIR) 374-200-90-00301. ,

The entire calibration procedure for APRM E was conducted to ensure that no abnonnalities exist in the circultry, not seemingly related to the neutron flux trip circuit. -This c.alibration procedure

- completely checks the drawer, power supply ripple and regulation, and all trip functions and indications. No out of callbration conditions were found.

CRD 26-47 (S/W A994) has already been scheduled for replacement during LaSalle's Unit 2 third refueling outage (L2R03) which is scheduled to begin in March of 1990. Work Request L94782 was written previously due to the degraded stall flows and high tangeratures ou CRD 26-47.

Action item Record 374-200-90 4 0302 has been generated to track the inspection and detennination of the failure mode for CR0 26-47. This action will be coupleted once the CRD is removed during L2R03.

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Fore Rev 2.0 LICEaKEE EVENT REPORT (LER) TEXT CONTINUATION LER INSER (6) Pane (3) .]

DOCKET NU SER (2)

FACillif NRME (1) ' Year // Sequential //j/ Revision fg//

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/// thsiber i m 11e County Station unit 2 015 I o I o 10131714 9Io - oIoI1 - oIo 01 5 0F 01 5 TEKT Energy Industry identification System (Ells) codes are identified in the text as (XXJ E. CORRECTIVE ACTIONS (Continued)

Furthennore, as control rod 26-47 may again drift from position "00" to position "02" after a scram -

reset, the shutdown margin analysis was revised to assisme that control rod 2647 is at "02" along with the highest worth rod being fully withdrawn, prior to the startup after the scr m. A caution card was i

placed on the Control Room bench board (scrm reset switch) to lainediately select and insert CRD 2647 when resetting a full scrm. This caution card will be removed after the drive replacement is coupleted.

F. ' PREVIOUS EVENTS LER Number Title 374/89-011-00 Spurious Reactor Protection System Actuation Due to Unknown Causes G. COMPONENT fAlLURE DATA N/A ,

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