ML20045D679

LER 93-002-01:on 930107,determined Potential for Premature Opening of Containment Bldg Spray Sump Isolation Valves. Caused by Failure to Adequately Recognize Potential Sys Interactions.Affected Procedures Revised

ML20045D679
Person / Time
Site:	Seabrook
Issue date:	06/21/1993
From:	Peschel J NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:
Shared Package
ML20045D677	List: ML20045D675 ML20045D679
References
LER-93-002, LER-93-2, NUDOCS 9306290292
Download: ML20045D679 (4)

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NRC FORM 366 U.S. NUCLEAR REGULATOR 7 LOMMISSION APPROVED BV OMB NO. 3150-0104

, (5-92) . EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY

" t "

LICENSEE EVENT REPORT (LER) g T 159. INF0R"^r0RWARD MM N S R BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714),

(See reverse for required number of digits / characters for each U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555 0001, AND TO THE block) PAPERWORK REDUCTION PROJECT (3150-0104),

OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

F ACILIIY NAME (1) DOCKET NUMBER (2) PAGE (3)

Seabrook Station 05000443 1 OF 4 IIILE (4)

Potential Premature Opening of CBS Sump Isolation Valves EVENT DAll (5) LER NUMBER (6) REPORT DATE (7) OTHER F ACILITILS INVOLVED (8)

SE AL R N MONTH DAY YEAR YEAR MONTH DAY YEAR 0500

^ "#"' U" 01 07 93 93 --002 -- 01 06 21 93 050 ()

OPERATING y IHIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMEN1S OF 10 CFR %: (Check one or more)

MODE (9) 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b)

POWER 20.405(a)(1)(1) 50.36(c)(1) X 50.73(a)(2)(v) 73.71(c) 100 LEVEL (10) 20.405(a)(1)(li) 50.36(c)(2) 50.73(a)(2)(vil) OTHER 20.405(a)(1)(iii) X 50.73(a)(2)(i) 50.73(a)(2)(viii) (Specify in Ab$ trac ow 20.405(a)(1)(iv) X 50.73(a)(2)(11) 50.73(a)(2)(viii) nl 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) NRC form 366A)

LICENSt[ CONTACI FOR THIS LFR (12)

NAME TELEPHONE NUMBER (Include Area Mr. James M. Peschel, Regulatory Compliance Manager Code)

(603)474-9521 ext. 3772 COMPLLIE ONE LINE F OR EACH COMPONENT F AIL URE DESCRIBED IN THIS RE PORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER R 0 E CAUSE SYSTEM COMPONENT MANUFACTURER R 0 E SUPPLEMENI AL REPORT EXPLCIED (14) EXPECTED MONTH DAY YEAR YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). X NO DATE (15)

ABSlRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On January 7, 1993, North Atlantic Operations personnel questioned whether performance of surveillance testing on the Refueling Water Storage Tank (RWST) [BP] level instrumentation, which tripped two level channels in Mode 1, was consistent with Technical Specifications. Evaluation has determined that the RWST level instruments were inoperable during performance of the surveillance testing, in addition, it has been determined that if a Safety Injection (SI) [JE] signal had occurred while the RWST level low-low bistables were tripped during surveillance testing, the Containment Building Spray (CBS) [BE] recirculation sump isolation valves would open earlier than previously analyzed. At that time it was postulated that the premature opening of the valves during a steam line break could potentially cause air binding of the CBS and Residual Heat Removal (RHR) [BP]

pumps. Therefore, the event was reported to the NRC on January 8, 1993 pursuant to 10CFR50.72(b)(2)(iii).

Engineering evaluation has determined that a large loss of Coolant Accident (LOCA) occurring while two RWST level low-low btstables were in the tripped condition could have resulted in containment temperatures and pressures, and peak cladding temperatures which may not have been bounded by the current analyses.

The root cause of the event is f ailure to adequately recognize potential system interactions during abnormal configurations occurring during surveillance testing.

Corrective action was to revise the af fected procedures to eliminate placing two RWST level channels in the tripped condition. In addition, North Atlantic will review surveillance procedures which provide a partial actuation of an ESF system to determine any potential safety implications. This event has been reviewed and discussed with the operating crews and other select North Atlantic personnel.

There were no adverse safety consequences as a result of this event.

9306290292 930621 PDR ADOCK 05000443 s PDR

NRC EORM 366A U.S. NUCLEAR REGULA10RY COMMISSION APPROVED BY OMB NO. 3150-0104

, (5-92) . EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST:

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TEXT CONTINUATION NUCLEAR REGULATORY COMMISSION WASHINGTON, DC 20555-0001 AND TO Tile PAPERWORK REDUCT]DN PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKE1 NUMBER LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVIS10 Seabrook Station "

05000443 93 01 2 OF 4

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1 Ext (lf more space is required, use addItlonaI coples of NRC form 366A) (17)

On January 7,1993, North Atlantic Operations personnel questioned whether performance of surveillance testing on the Refueling Water Storage Tank (RWST) [BP] level instrumentation, which tripped two level channels in Mode 1, was consistent with Technical Specifications. Evaluation has determined that the RWST level instruments were inoperable during performance of the surveillance testing. In addition, it was determined that if a Safety injection (SI) [JE] signal had occurred while the RWST level low-low bistables were tripped during surveillance testing, the Containment Building Spray (CBS) [BE] recirculation sump isolation valves woul! open earlier than previously analyzed. It was postulated that the premature opening of the valves during a steam line break could potentially cause air binding of the CBS and Residual Heat Removal (RHR) [BP] pumps. Therefore, the event was reported to the NRC on January 8,1993 pursuant to 10CFR50.72(b)(2)(iii). This event is now being reported pursuant to 10CFR50.73(a)(2)(i), (ii), and (v).

Backcround Information At Seabrook Station, the Emergency Core Cooling System (ECCS) takes a suction from either the RWST or the CBS sump. The RWST is utilized as the borated water supply during the injection phase of an accident.

When the supply of borated water in the RWST has been injected a transfer to the CBS sump is initiated.

The transfer is accomplished with both automatic and manual actions.

The RWST level instrumentation has four loops which provide an input into the two out of four logic required to generate a RWST level low-low signal. When actuated, this signal in combination with an SI signal will automatically open the Train A and B CBS sump isolation valves.

Technical Specification Table 4.3-2 specifies that a Channel Calibration be performed on the RWST level low-low coincident with a SI logie at least once per eighteen months. This surveillance also verifies that when two out of four RWST level channels are at their low-low level setpoint that a RWST level low-low signal is generated. As previously stated, this signal in combination with an SI signal will automatically open both CBS sump isolation valves.

Event Description On January 7,1993, with the reactor at 100% power, a Channel Calibration was performed on the RWST level circuitry. While performing Surveillance Procedure IX1622.231, "L-930 Refueling Water Storage Tank Level Calibration", the Unit Shift Supervisor questioned the presence of the RWST level low-low alarm in Mode 1. This indicated that two RWST level channels were in a tripped condition. Instrumentation and Control Department supervision were contacted and the RWST level transmitters were returned to service from the tripped condition. It was later determined that even though the RWST level instruments were in their required safeguards condition the level channels were inoperable because they would permit the CBS sump suction valves to open at a level other than the RWST low-low level setpoint.

During a preliminary evaluation of the condition it was postulated that if an SI actuation due to a steam line break occurred simultaneously with two out of four RWST level channels being tripped that the Train A and Train B CBS sump isolation valves would automatically open with the potential to air bind the CBS and RHR pumps. This was determined to be a condition which alone may have prevented the fulfillment of the safety

NRC E0RM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92) . EXP!RES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST:

50.0 HRS. FORWARD COMMENTS REGARDING LICENSEE EVENT REPORT (LER) BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.

TEXT CONTINUATION NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER LER NUMBER (6) PAGE (3)

YEAR

, SEQUENllAL REVIS10 Seabrook Station "

05000443 93 01 3 OF 4

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TEXT (If more space is required, use additional copies of NRC form 366A) (17) function of structures or systems that are needed to remove residual heat and to mitigate the consequences of an accident. Therefore, a non-emergency four hour report was made to the NRC pursuant to 10CFR50.72(b)(2)(iii).

Safety Conscouences There were no adverse safety consequences as a result of the event.

Adverse safety consequences could have resulted if a large LOCA occurred concurrent with the surveillance testing which placed two RWST level channels in trip. The potential safety implications of this scenario are addressed below.

Potential Safety Imolications Further evaluation determined that air binding of these pumps would not occur during a postulated main steam line break event as the pumps would continue to take suction from the RWST due to the elevation head from the RWST exceeding the containment pressure during the relatively short duration of this event.

Two scenarios were evaluated ta determine the effect of a postulated LOCA occuring while the RWST level channels were tripped. The first scenario assumes that the ECCS and spray flows collect in the rceirculation sump. The second scenario considers the Seabrook Station design which traps 17,000 cubic feet of fluid.

This fluid is not available to the recirculation sump.

Engineering evaluation has determined that approximately eight minutes into a postulated large LOCA, specifically a double ended cold leg guillotine break, occurring with two RWST level channels tripped, the pressure from the CBS sump boosted by containment pressure would exceed the pressure from the RWST.

This would result in the CBS pumps taking suction from the CBS sump prior to the normal time for transfer to the recirculation mode of emergency core cooling. The early suction from the CIlS sump would result in an elevated containment spray temperature early in the accident sequence and reduced spray effectiveness for pressure control and heat removal. This could have resulted in containment temperatures and pressures, and peak cladding temperatures which may not have been bounded by the current analyses.

In addition, the premature transfer to the CBS sump would also result in the RHR pumps taking suction from the CBS sump earlier than analyzed. This would increase the temperature of the low head injection fluid.

The consequences of the increase fluid temperature is a change to the reactor vessel reflood rate and a decrease of the decay heat removal effectiveness. The ultimate effect of the increased injection fluid temperature could be an increase in peak clad temperature.

If the ECCS and containment spray flows are assumed to initially fill the volumes unavailable to the recirculation sumps, then in a postulated double ended pump suction guillotine break the containment pressure alone would exceed the RWST head at the RHR suction check valve approximately 6 minutes into l

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! NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION' APPROVED BY OM8 NO. 3150-0104 l '

( 5*-92 ) . EXPIRES 5/31/95 1 .

ESTIMATED BURDEN PER RESPGCSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST:

50.0 HRS. FORWARD COMMENTS REGARDING L'ICENSEE EVEN"' REPORT (LER) BURDEN ESTIMATE TO THE INFORMATION AND TEXT CONTINUATION RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.

NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE- 0F MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.  ;

FACILITY NAME (1) DOCKET NUMBER LER NUMBER (6) PAGE (3) l YEAR SEQUENTIAL REVIS10  ;

Seabrook Station 05000443 " - 4 OF 4 93 --

0 2 --

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TEK1 (11 more space is required, use additional copies of NHC Form 366A) (17) the event. This could result in inadequate sump levels for the operation of the CBS and RHR pumps. j Suction vortexing or air binding of the pumps could occur, potentially resulting in a loss of CBS capability i and RHR system operation, which could prevent the establishment of long term core cooling. In this scenario, the potential exists for containment bypass with significant offsite doses.

The surveillance which trips the RWST level channels has been performed five times in MODE 1.

Root Cause The root cause of the event is failure to adequately recognize potential system interactions resulting from an abnormal configuration during surveillance testing. The event represents a situation where the consequences of placing individual components in their safeguards configuration were not adequately reviewed in terms of the integrated system performance under postulated accident conditions.

Corrective Action

1. Station Procedures have been revised to climinate placing the plant in a configuration with two out of four RWST level channels in the tripped condition in Modes 1-4

2. North Atlantic will review surveillance procedures which provide a partial ESF actuation, similar to IX1622.231, to determine if any similiar potential safety implications exist. The review will include partial ESF actuations and integrated system operation under normal and accident conditions.

3. North Atlantic Operations management has reviewed this event, to include the placement of components in their safeguards configuration, with the operating crews.

4. Select personnel from Licensing, Operations, Technical Support, and Engineering have participated in a Station Operating Experience Review seminar.

Previous Occurrences This is the first event of this type at Seabrook Station.

At the time of the event the plant was in Mode 1 at 100% power.

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