ML20038A907

Evaluation of Steam Line Break Consequences Associated W/Removal of Rupture Matrix Signals from Emergency Feedwater Valves.

ML20038A907
Person / Time
Site:	Crystal River
Issue date:	05/23/1980
From:	FLORIDA POWER CORP.
To:
Shared Package
ML20038A906	List: ML20038A905 ML20038A907
References
NUDOCS 8111240426
Download: ML20038A907 (4)

Text

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- I Rev.1 6/27/80 Prepared By Q [ b

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Rev.iewed By L Approved By _ M[ #JO F

EVALUATI0tl 0F.

STEAM LItlE BREAK C0tiSEQUE!1CES

ASSOCIATED WITH REMOVAL. OF RUPTURE MATRIX SIGilALS FROM EMERGEftCY FEED'.lATER VALVES -

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Prepared For Florida Power Corporation CR-3 Pro.iect May 23, 1980 A

Prepared By [b Reviewed By h Reviewed By fu__

l Approved By

• 1 8111240426 8111 i7' PDR ADOCK 05000302 P PDR -

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,. III. Containment Overpressurization Analysis

( An evaluation of the reactor building pressure response to a double-ended steam line break followed by unmitigated emergency feedwater flow has been made. The effect of the mass and energy releases on the reactor building pressure has been evaluated using the methodology described in Section 14.2.2.1.5 of the Crystal River III FSAR.

It was assumed that the containment is an adiabatic closed system. Steam gene ator blowdown is assumed to be instantaneously released to the containment as saturated steam. Emergency feedwater is added to the blowdown as a function of time (880 gpm @ 1170 BTU /lba)'. Static calculations were then performed at specific times in order to determine the duration of unmitigated EFW ficw that would be required to exceed the reactor building design pressure of 55 psig. It has been determined that approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of continuous EFW flow released to the containment as saturated steam in combination with normal spray action would be required to raise the containment pressure to it's design limit.

The reactor building spray system is normally used for heat removal and (j pressure suppression in an' elevated containment environment. Using reacto-building sprays of 3000 gpm provides significant pressure suppression. A curve of reactcr building pressure versus time after EFW initiation, with use of sprays is shown in Figure 2. Sufficieat time exists to allow for operator action to evaluate the steam generator situation and terminate EFW to the affected steam generator.

Containment evaluations have been performed for several B&W plants consider-ing unmitigated secondary system releases either from main or emergency systems. It can be concluded from these studies that main feedwater isolation following SLB is essential to prevent overpressurization in relatively short periods of time. Releases typical of emergency feedwater flowrates can be handled by building sprays for extended periods of time.

Rev. 1 6/27/80

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. 3 Figure 2 CRYSTALRihER111 REACTOR BUILOING PRESSURE FOLLOWING EFW AND SPRAY ACTUATION 80-70 -

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e 50 -

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( E EFW AND SPRAYS V 2 5

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30 t

20 e e 3 , i 8 0 10 20 30 40 50 60 e.

Time, minutes REV-1 0-27-80 9

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.o j V. Conclusions and Recommendations Based on the foregoing evaluations and analysis, the following conclusions were reached:

1) Operator action within one hour to isolate EFW will prevent exceeding the containment design pressure following a design basis steam line break. While spray syste'm operation is assumed, passive heat sinks are conservatively neglected.

2) The probability of a steam line break is low (% 10 -4 per reactor year).

The probability of an SLB and a stuck control rod (a necessary pre-requisite for a return to criticality) is conservatively estimated to be less than 2.75 x 10-7 per reactor year. Furthermore, though not specifically analyzed, the continued EFW flow case would be expected to show results similar to the FSAR analysis results for continued MPd flow.

3) Based on 1) and 2) above, we conclude that the incremental risk of containment overpressurization or return to criticality following a double-ended SLB and no automatic isolation of EFW to the affected stea.n generator is negligible.

4) While not a specific acceptance requirement for the FSAR analysis, contain-ment temperature associated with the above pressure analysis was reviewed.

The saturated steam pressure corresponding to the containment design temperature of 281 F is 50 psia. As can be seen from Figure 2, this corresponds to ~approximately 35 minutes, again demonstrating that adequate operator action time is available for isolation of EPJ to the affected genera tor.

5) The proposed change (elimination of the rupture matrix signals from the EFW valves) is consistent with the flRC recommendat ions (flVREG-0667) l and similar analysis performed for other projects.

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6) In light of the expected significant improvenents in secondary cooling reliability which can be realized and the corresponding minimal risk of containment overpressurization or core damage which would result, the proposed change should be implemented.

Rev. 1 6/27/80 L