Event Notification Report for June 07, 2001

POTENTIAL NON-CONSERVATISM IN MAIN STEAM LINE BREAK ACCIDENT IDENTIFIED DURING REACTOR POWER UPDATE ANALYSES

"During preparation of a new analyses for reactor power uprate, a potential non-conservatism was discovered regarding the main steam line break accident with an assumed single failure of a main feedwater regulating valve. This non-conservatism could negatively impact the resultant peak containment pressure that would be reached in the event of an accident. The current Point Beach accident analysis does not appear to properly account for the volume of high temperature feedwater that would be released inside containment during a main steam line break accident. It does not appear to correctly model failure of the feedwater regulating valve to close.

"Based on the revised analyses performed by the reactor vendor (Westinghouse) at the uprated power conditions, peak containment pressure could reach 64.2 psig. No revised analysis data is yet available for the currently licensed power level; however, the uprated analysis bounds the current plant design. The 64.2 psig value exceeds the containment design pressure of 60 psig.

"The Bases for Technical Specification 1 5.3.6.A.1 .a states, The safety design basis for the containment is that the containment must withstand the pressures and temperatures of the design basis LOCA without exceeding the design leakage rate. The design allowable leakage rate La is 0.4% of containment air weight per day at 60 psig . Containment operability is maintained by limiting the overall containment leakage rate to within the design allowable leakage rage (La).

"FSAR section 5.1 documents that the containment can withstand pressure loadings at least 50% greater than those calculated for the postulated loss-of-coolant accident alone, and 25% greater than those calculated for the postulated loss-of-coolant accident with a coincident design earthquake (FSAR 14.3.4 states that containment pressure will peak between 52 and 54 psig).

"Since the revised analysis is only for uprated power conditions, it does not exactly predict containment pressure results at the current licensed power level. However, the new analysis calls the results of the current analysis into question. Therefore, this report is being conservatively made to document this concern. The containment has been determined to be operable per the guidelines of Generic Letter 91-18. Mitigating circumstances include the fact that calculations in the FSAR document that the containment is capable of withstanding the higher pressure of the revised analysis.

"The containment will continue to perform its safety function. Any radiological release during a main steam line break accident is assumed to vent directly to the atmosphere (through the break), independent of containment integrity and is within current analyses."

Westinghouse is performing a reanalysis with corrective actions to be determined. The licensee informed the NRC resident inspector.

HPCI DECLARED INOPERABLE DUE TO A FAULTY FLOW INDICATING SWITCH

With DAEC operating in mode 1, at approximately 47% power, HPCI was declared inoperable due to a faulty flow indicating switch which is utilized to position the minimum flow valve. A 14-day LCO was entered per Technical Specification 3.5.1, condition F, at 2133 on 6/7/01. This condition is reportable under 50.72(b)(3)(v)(D) and 50.73(a)(2)(v)(D) as a failure of a single train of equipment required to mitigate the consequences of an accident.

While an in-plant operator was performing his rounds within the HPCI room, he discovered that flow indicating switch, FS2310, which should indicate total HPCI flow, indicated approximately -170 gpm rather than the expected zero gpm. The HPCI pump was not in operation. This flow switch is interlocked with the HPCI minimum flow valve, allowing it to open with total HPCI flow less than 300 gpm and HPCI discharge pressure greater than 125 psig. The flow switch would close the minimum flow valve once HPCI total flow reached 600 gpm. With the correct operation of the HPCI minimum flow valve in question, HPCI was declared inoperable.

It is believed that the cause of the faulty indication is air in the instrument sensing lines feeding the flow switch. Troubleshooting efforts will begin early tomorrow morning.

The NRC Resident Inspector was notified

HOO NOTE: see event #38029

* * * RETRACTED AT 1359 EDT ON 7/11/2001 BY JOHN KARRICK TO FANGIE JONES * * *

"This condition was reported (EN #38058, June 7, 2001) as 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to [D] mitigate the consequences of an accident.' NUREG-1022, Rev. 2 guidance for this reporting criterion essentially requires any unplanned HPCI LCO at BWRs to be reported as a loss of a single train safety system. Further review of this event indicates that, since only the flow switch for the HPCI minimum flow valve was inoperable, the minimum flow valve instrument Tech Spec could and should have been entered instead of the HPCI system LCO. The instrument spec. (TS 3.3.5.1, Condition E) provides a 7-day completion time to return the instrument to service. If the instrument is not restored in 7 days, then the 14-day HPCI system LCO would be entered. The instrument was returned to service the following day. In addition, an assessment of the as-found condition (irrespective of TS) concluded that the HPCI system was capable of performing its intended safety function with this switch in its as-found condition. Therefore, there was no event or condition that could have prevented safety function fulfillment and this event is being retracted. No other reporting criteria apply to this event."

The licensee notified the NRC Resident Inspector. The R3DO (David Hills) was notified.