ML20012B293

LER 90-002-00:on 900204,determined That Only Reactor Bldg Differential Pressure of 0.24-inch Water Vacuum Could Be Obtained,In Violation of Tech Spec Required 0.25 Inch.Caused by Testing Deficiency.Leak Paths sealed.W/900305 Ltr

ML20012B293
Person / Time
Site:	Quad Cities
Issue date:	03/05/1990
From:	Bax R, Schumacher D COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-002-05, RLB-90-072, NUDOCS 9003140141
Download: ML20012B293 (6)

Text

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i; a- Commomssalth Edison.

Quad Cities Nuclear Power Station i 22710 206 Avenue North L ooroova, Illinois $1242 Telephone 300/664 2241 RLB-90-072 March 5, 1990 U. S. Nuclear Regulatory Commission Document Control Desk Hashington, DC 20555

Reference:

Quad Cities Nuclear Power Station Docket Number 50-254, DPR-29. Unit One Enclosed is Licensee Event Report (LER)90-002, Revision 00, for Quad Cities Nuclear Power Station.

This report is submitted in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 50.73(a)(2)(1)(B): The licensee shall report any operation or condition prohibited by the plant's Technical Specifications.

Respectfully, COMMONNEALTH EDISON COMPANY QUAD CITIES NUCLEAR POWER STATION

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Station Manager RLB MJB/ekb Enclosure cc: R. Stols R..Higgins INPO Records Center NRC Region III f

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LICENSEE EVENT REPORT (LER) Fare key 2.0 Facility Name (1) Docket Number (2) Pane f3)

' nuad citiet unit One of El el of 01 fl El 4 1 ef C E Title (4) Inability to Maintain 0.25 Inches of Water Vacuum Ouring secondary Containment Test Due to T rature Effnett.

Event Date (E) LtR Nunter (6) Recort Date f71 Other Facilities Involved fR)

Month Day Year Year g/p Sequential //j/

//

ff Revision Month Day Year Facilits Namet Docket Numberf t)

/// Number /// Number Quad citiet 01 El 01 01 01 21 61 1

~ '"

el 2 OI 4 91 0 91 0 0 10 l2 0 10 0I3 DIE 91 0 01 E l o f 01 01 l l THIS REPORT IS $ubMITTED PUR$uANT TO THE REQu!REMENTS OF 10CFR OPWW (Check one or more of the f ollowine) fll) 4 20.402(b) 20.405(c) 50.73(a)(2)(tv) _._ 73.71(b)

PodER __ 20.40$(a)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

LEVEL _, 20.405(a)(1)(11) _ 50.36(c)(2) 50.73(a)(2)(v11) Other (Specify (101 0 9 1 20.405(a)(1)(tit) )L 50.73(a)(2)(1) 50.73(a)(2)(vitt)(A) in Abstract

//////////////////////////, 20.40$(a)(1)(iv) 50.73(a)(2)(11) 50.73(a)(2)(viti)(B) below and in

//////////////////////////' , 20.405(a)(1)(v) 50.73(a)(2)(111) 50.73(a)(2)(x) Text)

LItthitt CONTAtf FOR THIS LER f12)

Name T[LEPHONE NUMBf R AREA CODE

0. F. Sch-her. Technteal Staf f Ennineer. Ext. 2820 3 10 l9 61 El 4l -l fl 214l COMPLETE ONE LINT FOR [AtH COMP FAILURE DEstRIBt0 IN THIS REPORT (13)

CAuSE SYSTEM c0MPONENT MANuFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANuFAC. REPORTABLE TURER TO NPROS TURER TO NPR01 I l I l i I I N I l l l l 1 l 1 l ! I I I I I I I I I I l SUPPLEMENTAL REPORT trP[tTED (141 Expected t!Qalh I Day l Year submission lYet (If vet. enmelete txPttit0 tusMISSION DATE) K l NO ' (1 }

l ll lI ABSTRACT (Limit to 1400 spaces, i.e. approximately fif teen single-space typewritten lines) (16)

ABSTRACT:

At 0754 hours0.00873 days <br />0.209 hours <br />0.00125 weeks <br />2.86897e-4 months <br /> on February 4, 1990, Unit One was operating in the RUN mode at 91 percent rated core thermal power. A GSEP unusual event was entered due to the results obtained while performing QTS 160-5, Secondary Containment Capability Test. It was determined that a Reactor Building differential pressure of 0.24 inches of water vacuum could be attained; whereas, an average of 0.25 inches of water vacuum is required by Technical Specification 4.7.C.I.c.

Inspections of suspected leak paths were conducted and temporary seals were installed. By 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />, the reactor building differential pressure had reached 0.275 inches of water vacuum. Seven temporary seals were removed with little or no effect on the differential pressure. At 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> the differential pressure had reached 0.28 inches of water vacuum.

The cause of this event is attributed to temperature induced differences in the pressure gradients between the inside and outside of Secondary Containment, which were uncompensated for in the test procedure. Due to this effect, it is believed that Secondary Containment capability was intact throughout this event.

l At 1642 hours0.019 days <br />0.456 hours <br />0.00271 weeks <br />6.24781e-4 months <br />, on-site review 90-5 was performed to review the actions involved with the GSEP unusual event. At 1648 hours0.0191 days <br />0.458 hours <br />0.00272 weeks <br />6.27064e-4 months <br />, the GSEP unusual event was terminated.

This report is being submitted in accordance with 10 CFR 50.73(a)(2)(1)(B).

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i tittuttr tvtut arrant strai ttri enmitmuatinu rara may 2.0 f

FACILITY mAME (1) DOCKET WuMsER (t) Ltn utsa tt r n Pane in k Year seguential Revision name. r er j w cities unit one o I s I o I e I o i 21 51 4 910 - oIo12 - oI o el 2 or als j TEXT Energy Industry 16entification system (E!!s) codes are identified in the text as (xx] l PLANT AND SYSTEM IDENTIFICATION:

i General. Electric - Bolling Water Reactor - 2511 MWt rated core thermal power.

EVENT IDENTIFICATION: Inability to Maintain 0.25 Inches of Water Vacuum During q Secondary Containment Test Due to Temperature Effects.

l A. CONDITIONS PRIOR TO EVENT:

Unit: One Event Date: February 4, 1990 Event Time: 0754 Reactor Mode: 4 Mode Name: RUN Power Level: 91%  ;

i This report was initiated by Deviation Report D-4-1-90-016 RUN Mode (4) In this position the reactor system pressure is at or above 825 psig, ,

and the reactor protection system is energized, with APRM protection and RBM  ;

interlocks 1.1 service (excluding the 15% high flux scram).

B. DESCRIPTION OF EVENT: i At 0754 hours0.00873 days <br />0.209 hours <br />0.00125 weeks <br />2.86897e-4 months <br /> on February 4, 1990e Unit One was operating in the RUN mode at 91 percent of rated core thermal power, and Unit Two was in the SHUTDOWN mode for the beginning of a scheduled refueling outage. Technical Staff personnel were performing QTS 160-5, Secondary Containment [NH] Capability Test. It was  ;

determined that an average of 0.25 inches of water vacuum could not be maintained in the Reactor building as required by Technical Specifications.

l Initial test results indicated 0.24 inches of water vacuum could be attained in the Reactor Building [NH). Based on this result, Technical Specifications require that within twenty four hours, both units be placed in a condition that does not require Secondary Containment. Unit Two was in cold shutdown which satisfied the requirement, and at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />, a shutdown of Unit one reactor was initiated at a rate of 100 MWe per hour and a GSEP unusual Event was entered. ,

OTS 160-5 was initiated at 0351 hours0.00406 days <br />0.0975 hours <br />5.803571e-4 weeks <br />1.335555e-4 months <br /> on February 4, 1990. The wind speed was approximately 8.5 miles per hour and the direction was 210 degrees. The average of 4 differential pressure indicators located at the 696' elevation of the Reactor  :

Building showed 0.24 inches of water vacuum.

Between 0800 and 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />, inspections were made of suspected leak paths which Cere then temporarily repaired to determine their effect on Reactor building  ;

differential pressure. By 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />, the reactor building differential pressure had increased to 0.275 inches of water vacuum.

With Reactor building differential pressure at an acceptable value, 7 of the 11 temporary seals were removed one at a time in an attempt to quantify the leakage and it was concluded that these seals had had little or no effect on the building differential pressure. At 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, with the average differential pressure at 0.28 inches of water vacuum, the test was successfully completed.

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LifEMitt EVENT RfPORT fLER) T[KT fDMTIMHATIhN Form Rev t.o FACILITY NAME (1) 00cKti Nupett (2) Ltk munare f 61 Paae f21 Year // sequential /// Reviston

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/// u d ar med fit tes unit one o l's I o I o I e 121114 eIo . oIoIt . oI o el 3 or als TEXT Energy Industry 16enttf tcation system (E!!s) codes are 16entified in the text as [XXI At 1642 hours0.019 days <br />0.456 hours <br />0.00271 weeks <br />6.24781e-4 months <br /> on-site review 90-5 was performed to review the actions involved with  !

the GSEP unusual event. At 1648 hours0.0191 days <br />0.458 hours <br />0.00272 weeks <br />6.27064e-4 months <br />, the GF.P unusual event was terminated, the

• Unit One shutdown was stopped and the station began returning ventilation systems -

to normal operation.

C. APPARENT CAUSE OF EVENT.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(1)(B): The licensee shall report any operation or condition prohibited by the plant's -

Technical Specifications.

The primary cause of this event has been attributed to a testing deficiency.

Contributing to the event are a number of minor air leaks which were identified during the test. ,

The test did not encompass temperature induced differences in the pressure gradients between the inside and outside of Secondary Containment. Thus, this temperature effect resulted in an apparent, but most likely not an actual loss of Secondary Containment capability.

The effects of a temperature difference between Secondary Containment and atmosphere are outlined in the Nuclear Regulatory Commission's (NRC) Information Notice No. 88-76: Recent Olscovery of a Phenomenon Not Previously Considered in the Design of Secondary Containment Pressure Control. When the temperature of the ,

atmospher(. is lower than the interior of the building, the difference in air density will create a pressure differential which changes with elevation.

i The test procedure requires the measurement of the building to atmosphere differential pressure at the 696' elevation, near the roof line. When the building temperature is higher than the outside tempere.ture, the differential pressure .

measured at this elevation can be considerably less than the differential pressure l at ground level, 595' elevation. A measured differential pressure with respect to atmosphere at the 696' elevation is more conservative than a differential pressure I based on an overall average of Reactor Building elevations.

When the GSEP unusual event was declared at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />, the building wat more than I 50 degrees Fahrenheit (*F) warmer than atmosphere. At this time the measured J differential pressure was 0.24 inches of water vacuum at the 696' elevation.  !

Preliminary calculations performed by station Technical Staff indicated tnat the l average differential pressure of Secondary containment to atmosphere, compensated for temperature affects, was greater than 0.26 inches of water vacuum.

l Therefore, it is believed that Secondary Containment capability was intact throughout this event.

)

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i Lffluitt tVthT ktPQRT (LER) TEXT CONT 1kuATIDW Fora tev 2.o

? FACILITY NAME (1). D0cKET NuMetR (2) __ t[a uuMatt ist Pane 131 -

t Year / sequential Revision g//

/ Number //, Number

' m cities unit one o i s I o'l o I o 1 21 si 4 oIo - oIol2 - oI e of 4 or als ttxt Energy Inoustry Identification system (t!!s) codes are ioentified in the text as [XK) ,

Minor leaks were identified, temporarily sealed and left sealed until permanent repairs can be accomplished. These consisted of the Unit Two main steamline [ SEAL)

I seals and two feed water boots which were found to be degraded. These penetrations ,

connect the Main Steamline [SB) Isolation Valve (ISV) (MSIV) room, part of the Reactor building, and the High Pressure Heater (SB) Bay, part of the Turbine ,

o Building. Minor leaks were also left sealed on duct work access doors located in ,

the Drywell Torus Purge Fan [VB, FAN), Standby Gas Treatment, and MSIV room exhaust o systems [VA). These leaks are' thought to have had no significant impact on the '

l test results.

D.- SAFETY ANALYSIS OF EVENT:

L Technical Specification 4.7.C.I.c states Secondary containment capability to maintain an average 0.25 inches of water vacuum under calm wind conditions (2<uc5 cph) with a Standby Gas [BH) filter [FLT) train flow rate of not more than 4000 cfm shall be demonstrated at each refueling outage prior to refueling. Surveillance '

procedure QTS 160-5 is used to verify this requirement.

Technical Specification 3.7.C.1 states that in the event that Secondary Containment  !

integrity cannot be maintained, procedures shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to .

place the units in a condition that does.not require Secondary Containment. These '

conditions require both units be subtritical, reactor water temperature be below 1

212*F and the reactor coolant systems vented, no activity be performed which can I reduce the shutdown margin below specified, and no movement of the fuel cask or

• irradiated fuel in the Reactor Building.

This event was of minimal safety significance because preliminary analysis of this event indicates that Secondary Containment was within the required Technical Specification limits at all times. The neasurement of the building differential  ;

pressure at the high elevation of 696' is conservative. FSAR section 5.3.5 states that the reactor building to outside atmosphere average negative pressure be corrected for zero wind end zero differential temperature conditions. Additional '

conservatism was introduced by not correcting for zero differential temperature condition. Since the average differential pressure was 0.28 inches of water vacuum ,

at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, there was no loss in capability to perform its design safety function of limiting a radiation release during a design basis accident.

E. CORRECTIVE ACTIONS:

f The minor leakage paths were identified, temporarily sealed, reviewed, and approved  :

as satisfactory. They will be inspected on a weekly basis per on-site review 90-5 until permanent repairs are made.

The temperature effect phenomenon on Secondary containment will be further investigated under Action Item Record (AIR) 4-90-04. The Secondary containment ,

procedure. QTS 160-5, will be reviewed and revised to eliminate any deficiencies and incorporate the results of the temperature effect into this procedure (NTS 2542009001601).

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U t.1frutfr'EVtNT ttPORT fLER) TEYT CDNYtMAT10N Farm tav 2.0 [

FACILITY hAME (1) DOCKET NUMBER (2) Ltk NLSehrR fE) Paam (1) {

Year //

/ppjl ll seguential u m ar j/

/jl/

fjl Revision u'r I

. ' " etttet unit ans als!'IoIe121sia e eIo - eie12 . oI e el s or als TK5T Energy IndusVry toontification system (t!!s) cones are identified in the text as [xx]

'F. PREVIOUS EVENTS: ..

I There have been no previous events involving the failure of' Secondary Containment I

i. testing at this station. As this event was no a component failure, a Nuclear Plant .

Reliability Data System (NPRDS) search was not performed.

l G. COMPONENT FAILURE DATA: i This event did'not involve any component failure. '

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