ML19263E799

LER 79-017/01T-2 on 790330:util Notified by Westinghouse of Single Dropped Rod W/Potential for Calculated DNB Ratios Lower than Reported to NRC for Floating Nuclear Plants. Caused by Incorrect Measurement During Rod Drop Events

ML19263E799
Person / Time
Site:	Farley
Issue date:	06/18/1979
From:	Hairston W ALABAMA POWER CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML19263E798	List: ML19263E797 ML19263E799
References
LER-79-017-01T, LER-79-17-1T, NUDOCS 7906250292
Download: ML19263E799 (3)

Text

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EVENT DESCHIPTION AND PROS ABLE CONSEQUENCES h'd oi2 l WestinRhouse notified Alabama Power Co. that a review of safety analysis methodologv l o 3 l for the single dropped rod ~ indicated a potential for that event to lead to calculatedI oi.  !. DNB ratios lower than reported to the NRC for the FNP class of plant. The impact of l o s I this inconsistency is br j ved to ba minimal in that there are several mitigating l io o I effects not credited in the calculational method that significantly reduce the conse- !

O 7 l Quences of thin transient (see attachment). Engineering review of this notification [.

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3/ m 40 46 43 44 47 CAUSE CESCRIPTION AND CCRRECTIVE ACTIONS Q7) ilo IThis potential inconsistenev arose from two sources (a) the existing rod controller 1 W lcan potentially incorrectiv measure the core averaee oower level durine certain rod  !

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ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT DOCKET NO. 50-348 ATTACHMENT TO LER 79-017/01T-2 Facility: Joseph M. Farley Unit 1 Report Date: 6/1 8/ 79 Event Date: 3/30/79 Identification of Event Westinghouse notifi ation was received of a safety analysis methodology inconsistency in the NRC reviewed single dropped rod analysis.

Conditions Prior to Event The unit was in mode 6 at the time of notification.

Description of Event Westinghouse notified Alabama Power Company that a review of safety analysis methodology for the single dropped rod indicated a potential for that event to lead to calculated DNB ratios lower than reported to the NRC for the FNP class of plant. Engineering evaluation of this notification resulted in a determination on 3/30/79 that this inconsistency is potentially reportable under Tech. Spec. 6.9.1.8(h).

Designation of Apparent Cause This potential inconsistency arose from two sources:

a. The existing rod controller can potentially incorrectly msasure tne core average power level during certain rod drop events.

b. Deviations between rod control settings actually used in the field and those assumed at the time of the safety analysis.

Analysis of Event The ispact of this inconsistency is believed to be minimal in that there are several mitigating effects not credited in the calculational method that significantly reduce the consequences of this transient.

Among these effects are:

1. To meet the F limits required for LOCA and reduce burnup shadowing the 9 control rods during normal operation are typically inserted less than 5 to 10%. This corresponds to approximately 100 pcm of reactivity. The dropped rod that is assumed has a worth that is also typically 100 pcm. Thus, the rod controller, by withdrawing the control bank, can restore full power but generally cannot result in a power overshoot.

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79-017/01T-2

2. The dropped rod assumed is the most limiting rod in terms cf the resulting increase in F dropped, would result in muhN. Theincreases lower majority in of Frods, if aE'

3. Our analysis assuaes that the negative flux trip does not occur during this trataient. However, the available plant data indicate that most single dropped rods result in a negative flux rate trip. In fact, these rods that provide the limiting F

AH values are also most likely to provide a reactor trip.

4. The analysis presented in the FSAR assumed the bounding conservative reactivity coefficients allowed by the Teca.

Spec. The actual reactivity coefficients in the plant are significantly less limiting. The use of the actual moderator and doppler coefficients would reduce the power overshoot.

5. The FSAR analysis did not assume the operation of the overpower rod block Secause it is controt grade equipment. This block is expected to be in operation and would terminate rod motion when the power increases to 103% of nominal. This effect greatly reduces the potential tor an excessive power overshoot.

6. Improved safety analyses methods reviewed asd approved by the NRC on the D. C. Cook Unit 2 application (use of statistical DNB and the WRB-1 correlation), demonstrate the existence of significant margins compared to the margins shown in the original safety analyses for affected plants. Recognition of this margin, as well as other conservat've features of our overall safety analysis methodology, can provide high assurance that the single rod drop event does not, in fact, violate the accepted limiting DNB ratio.

7. The rod control system limits the amount of power overshoot during the red drop transient. For those plants where this effect is significant, minor modifications to the settings will further reduce the magnitude of power overshoot. Recog-nizing the importance of the rod control system performance during this transient, it is prudent to provide means in the form of procedures to ensure that the rod control system performs as design d.

Effect on Plant This occurreace had no effect on plant operation.

Corrective Action In accordance with Westinghouse recommendations, the Poe:r Range Nuclear Instrumentation negative rate trip setpoint will be changed from 1 5% of rated thermal power with a time constant > 2 seconds to 13% of rated thermal power with a time constant of > 1 second.

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, 79-017/01T-2 Also, the Power Range Nuclear Instrumentation positive rate trip setpoint will be changed from 5 5% of rated thermal power with a time constant 1 2 seconds to 1 5% of rated thermal power with a time constant 1 I second.

This setpoint change will protect the plant against single red drop accidents thus precluding possible DNBR violation subsequent to a rod drop. The revised setpoints will be incorporated in plauc procedures upon NRC approval of the associated Technical Specification change.

Failure Data None 2215 057'