ML20046C129

LER 93-007-00:on 930701,determined That Unit 1 Ssf Rc Makeup Sys Inoperable in Past Due to Design Deficiency.Operations Procedures Revised to Reflect Newly Calculated Operating Limits for Rc Makeup Pump,Rcps & RCS.W/930802 Ltr

ML20046C129
Person / Time
Site:	Oconee
Issue date:	08/02/1993
From:	Benesole S, Hampton J DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-007-02, LER-93-7-2, NUDOCS 9308090228
Download: ML20046C129 (7)

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.~ Duke 1%uttCompany _ '1 W lhwron Oconee NudearSite ncelyesident P.O.BoxI439 (803)Si&3199 Ollice

- Seneca,5C29679 (803M53564 Far . >

i DUKEPOWER .

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August 2, 1993 i

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

Subject:

Oconee Nuclear Station  ;

Docket Nos. 50-269, -270, -287 -

LER 269/93-07 Gentlemen Pursuant to 10 CFR 50.73 Sections (a)(1)'and (d), attached is Licensee Event Report (LER).269/93-07, concerning the technical inoperability of.

the alternate Reactor Coolant Makeup System.

This report is being submitted in accordance with 10 CFR 50.73 (a)(2)(1)(B). This event is considered to be of no significance with respect to the health and safety of the public.

Very truly yours, e  %'W A

J. W. Hampton ,

Vice President

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Attachment xc: Mr. S. D. Ebneter INPO Records Center Regional Administrator, Region II Suite 1500 U.S. Nuclear Regulatory Commission 1100 circle 75 Parkway 101 Marietta St., NW, Suite 2900 Atlanta, Georgia 30339 Atlanta, Georgia 30323 '

Mr. L. A. Wiens Mr. P. E. Harmon Office of Nuclear Reactor Regulation NRC Resident Inspector-U.S. Nuclear Regulatory Commission Oconee Nuclear Site Washington, DC 20555 c..

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Oconee Nuclear Station, Unit One 05000 269 1OF6 nTLEie Design Deficiency Results In The Technical Inoperability Of The Alternate Reactor Coolant Makeup System EVENT DATE (5) LER NUMBER (6) REPORT NUMBER m i OTHER FACILITIES INVOLVED (8) i g gggg + ACiUT't NAME DOCKET NUMBER WH DAY YEAR YEAR " * * ^"

NUMBEA NUMBER 05000

• ACsWTY NAME DOCKET NUMBER

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07 01 93 93 07 00 08 02 93 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 6: (Check orie or mores (11)

MODE (9) N 20 402ib) 20 405(c) 50.73(aH2)(iv) 73 71(: j POWER j 20 405!aH1)td 50.36(cH1) 50.73(aH2)(v) 73.71(c)

LEVEL (10) 100l 20.405:aH11ta) 50 36(cH2) 50.73!aH2Hvid OTHER 20.405faH1)oio x 50 73(aH2H4 B) 50.73(aH2HviiiHA) sP"W""

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LICENSEE CONTACT FOR THIS LER (12)

NAvt! TELEPHONE NUMBER tinciuoe Area Gocal S. C. Benesole, Safety Review Manager (803) 885-3518 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTUREA y CAUSE SYSTEM COMPONENT MANUFACTURER 3

SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED WM DAY YEAR l

vEs SUBMISSION N yn. conce EXPECTED $UBMISSON DAtn I DATE ( 1 ABSTRACT (Limit tD 1400 spaces. i e.. approxirnately 15 single spacea typewntten lines) (16) i In May 1992, Oconee Engineering (OE) initiated an evaluation of the parameters that affect the Reactor Coolant Makeup (RCMU) system operability. The evaluation was initiated after an excessive Reactor Coolant (RC) pump seal leakage event on Unit 1 (LER 269/92-09).

Specifically, the evaluation was to determine the adequacy of the RCMU system to supply the RC pump seals during a Standby Shutdown Facility (SSF) event. On July 1, 1993, with Unit 1 at 100% full power. OE determined that the Unit 1 SSF RCMU system had been inoperable in the past. The Unit 1 RC pump seal leakage rates have occasionally exceeded the newly established maximum allowed seal leakago rates. The root cause of the SSF RCMU system j inoperability is a Design Deficiency, functional design deficiency,  :

mechanical. Corrective actions included completing the design calculation '

for allowable RC pump seal leakage rates, revising appropriate procedures and initiating modifications to install higher accuracy flow gauges.

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SEQUENTA. REVISION NUMBER NUMBEM 05000 269 2OF6 Oconee Nuclear Station, Unit One 93 -

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MOVENT.A REV6cN NUMBER NUMBER 05000 269 3 OF 6 Oconee Nuclear Station, Unit One 93 -

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fact that during a Standby Shutdown Facility (SSF) event, decay heat from the spent fuel will not be removed by the Spent Fuel Pool Cooling system IEIIS:DA1. Therefore, the temperature of the water contained in the spent fuel pool will rise until the pool begins to boil at 212 F. The temperature of the spent fuel pool water will increase to a maximum of 220 F after passing through the SSF RCMU pump. The 220 F temperature is much hotter than the normal High Pressure Injection (HPI) [EIIS:BQ1 seal injection water temperature. Also, Unit 1 RC pump seal leakage rates increase as seal injection water temperatures increase. Therefore, seal leakage rates during an SSF event could exceed the SSF RCMU system's ability to provide seal injection flow if RC pump seal leakage rates during normal operation are too high. The OE preliminary evaluation of the l Westinghouse analysis results revealed that the seal leakage limit could be as low as 3.4 gpm based on a minimum of 26 gpm RCMU pump flow rate. Based on the preliminary results. Oconee Systems Engineering (OSE) began monitoring RC pump seal leakage to ensure the limit would not be exceeded before the evaluation was completed.

On June 1, 1993, prior to completing the evaluation, OSE discovered that the 1B1 RC pump seal leakoff was exceeding the recently established conservative limit of 3.4 gpm. The 1B1 RC pump seal leakoff was fluctuating between 3.45 and 3.50 gpm (with no adjustment for instrument error). OSE initiated a PIP report to document this problem.

In response to the PIP report, the OE evaluation concluded that 3.8 gpm RC pump seal leakage was acceptable if the RCMU pump delivers a minimum of 27 gpm. This was based on a conservntive flow instrument error of +/- 2 gpm.

Unit 1 RC pump seal and RCMU pump performance data and calculations concluded on June 3, 1993, that the actual seal leakage rates were acceptable.

On June 17, 1993, the calculations of the maximum allowed RC pump seal leakage rates and maximum allowed total combined RC system leakage for all I three units were completed and approved. This calculation incorporated a l

.7 gpm allowance for RCMU pump flow instrument error. Therefore, a minimum required RCMU pump flow rate of 28.3 gpm can be achieved. For Unit 1.

specifically, the maximum allowed RC pump seal Icakage rates based on providing a minimum of 28.3 gpm to the RC pump scair are: 1A1 RC pump -

4.7 gpm. 1A2 RC pump - 4.5 gpm. 1B1 RC pump - 4.2 gpm, and 1B2 RC pump -

4.7 gpm. The maximum allowed total combined RC system leakage rate is 16.8 gpm for the Unit 1 SSF RCMU system to be considered operable. The RCMU system was considered to be presently operable since the leakage rates had I not exceeded the calculated rates determined in the newl, approved engineering calculations.

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00 rEn on aa, a nwa. a. eaae.na coca.s a wC *m s'n An investigation was initiated to determine if the RC pump seal leakage had exceeded the new limits in the past. On July 1, 1993, it was identified that on one occasion, in November 1988, the seal leakage limits had exceeded the SSF Technical Specification Limiting Condition for Operation.

CONCLUSI0t(S The root cause of this event is a functional mechanical design deficiency, which occurred during the original design of the Reactor Coolant Makeup (RCMU) system in 1980-1981.

The original design calculations had not properly taken into accuwd the affect of spent fuel pool water temperature increases on the Reactor Coolant (RC) pump seal leakage. The design processes have been revised since the construction of the Standby Shutdown Facility (SSF). Guidelines for the calculations performed currently require a review of Quality Assurance Requirements for the Design of Nuclear Power Plants (ANSI N45.2.11) as an aid in the inclusion of all the applicable criteria. Duke Power has also established the Design Basis Document process which is intended, in part, to identify this type of oversight. The-corrective action from the May 1992 event (LER 269/92-09) led to the discovery of this problem. A corrective action from a previous event (LER 269/93-03) was to complete the Design Basis Document for the SSF.

This event is considered recurring. LER 269/91-12 identified a functional design deficiency related to a low setpoint on a RCMU pump relief valve.

LER 269/93-03 identified the technical inoperability of the RCMU system due to excessive nitrogen pressures in the RCMU pump suction stabilizer bladder. Because the RC pump seal leakage design deficiency has existed since the original operability of the SSF in 1980-1981, and since there has been no reason to reanalyze the RC pump seal leakage eniculations until the event in May 1992 (LER 269/92-09), no corrective action from these previously discovered events could have prevented it.

There were no personnel injuries, releases of radioactive materials, or NPRDS reportable equipment failures associated with this event.

CORRECTIVE ACTIONS Immediate None

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Subsequent

1. Revised operations procedures to reflect the newly calculated operating limits for the Reactor Coolant Makeup pump, Reactor Coolant pumps and Reactor Coolant system.

Planned

1. Change the present Reactor Coolant Makeup Pump flow gauges to gauges having a higher accuracy.

SAFETY ANALYSIS The High Pressure Injection (HPI) and Component Cooling (CC) [EIIS CC}  ;

systems provide cooling flows to the Reactor Coolant (RC) pump seals during normal plant operation. If these systems are unable to provide seal cooling, the Reactor Coolant Makeup (RCHU) system can be used to provide RC '

pump seal cooling, in addition to replenishing the RC system to offset seal leakage and RC system shrinkage during cooldown to hot shutdown. Each RC pump contains approximately 55 gallons of relatively cool (130-150 F.)

water between the RC pump seals and the RC systeE.. The cool water acts as a buffer between the hot RC system water and the RC pump seals. If this cool water leaks off before seal flow can be reestabliehed, the RC pump seals may degrade and the seal leakage rates may increase.

The seal leakage limit for the 1B1 RC pump was established at 4.2 gpm for a RCMU pump flow rate of 28.3 gpm after completing the design calculation.

The design calculation completed in June 1993 specifically identified the maximum allowable leakage for each pump. Previous leakage had exceeded the maximum allowable leakage at various times but not for an extended period.

The longest time that the leakage had been excessive was eight days, from November 16, 1988 to November 23, 1988. The leakage rate removed the assurance that the RC pump seals would not fail before the RCMU flow could be established. If the normal RC pump seal cooling systems had failed, then the degradation of the RC pump seals was possible. The resulting unisolable leakage from the RC system would be greater than the makeup capacity of the RCMU pump, resulting in a RC pump seal loss of coolant accident (LOCA). However, the probability that all normal seal cooling would have been lost during this time has been evaluated and found to be small (on the order of 1.0E-5 or lower). Thus, it was highly improbable that the SSF would have been necessary to provide RC pump seal cooling during the time that excessive seal leakage was present. 1 I

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NUMBER NUuBER 05000 269 6OF6 Oconee Nuclear Station, Unit One 93 -

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00 ru,m m ace w .aa..nm,.am awcFamaem on The Oconee Final Safety Analysis Report (FSAR) analyzes LOCA events, for a spectrum of break sizes that envelope RC pump seal LOCA's. The FSAR analyses demonstrato that the core will remain covered and radiological releases will remain within 10CFR100 limits, for seal LOCAs with HPI safety injection. RC pump seal LOCA events without HPI safety injection are not analyzed in the FSAR, because no plausibic single failure would fail the

• HPI and CC systems. However, this type of accident has been analyzed in support of safety evaluations for a station blackout (SBO). For a SB0 of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> duration with a postulated 25 gpm seal leakage per RC pump, the core will remain covered. With the core remaining covered, the radiological consequences of a RC pump seal LOCA are expected to be bounded by the FSAR Chapter 15 LOCA analyses.

The health and safety of the public were not compromised by this event.

Also, this event did not result in the release of any radioactive materials, radiation exposures or personnel injuries.

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