ML20214E076

Auxiliary Feedwater Calculation Summary

ML20214E076
Person / Time
Site:	Davis Besse
Issue date:	11/17/1986
From:	Carlton J, Domaleski E BABCOCK & WILCOX CO.
To:
Shared Package
ML20214D825	List: ML20214D821 ML20214E040 ML20214E076
References
77-1165893-01, 77-1165893-1, TAC-63772, NUDOCS 8611240344
Download: ML20214E076 (3)

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Page 1 AUXILIARY FEEDWATER CALCULATION

SUMMARY

77-1165893-01 for Toledo Edison Prepared by Babcock & Wilcox Nuclear Power Division 3315 Old Forest Road P.O. Box 10935 Lynchburg, Virginia 24506-0935 l

l Approved by _

E. J. Donaleski Reviewed by y' J. D. Carlton 8611240344 861117 PDR ADOCK 05000346 p PDR

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's B&W document $86-1158447-00 contains the results of several loss of feedwater cases which were run for Toledo Edison. For establishing acceptable emergency feedwater flow case 3 can be l

used as a basis. The analytic assumptions of interest for Case 3 are as follows o Initial power -- 100 %FP o Decay heat -- 1.0 times the 1979 ANS 5.1 decay heat curve o RC pumps operating o Reactor trip on high RC pressure (setpoint of 2300 psig) o AFW delivery to the steam generator 120 seconds after reaching a collapsed liquid level of 20" in the steam generator A flow rate of 650 gym will match decay heat and RC pump heat at a point less than 100 seconds following trip. Assuming that a total of 650 gym flow enters the steam generators 100 seconds after trip (this is equivalent to approximately 92 reconds after the 20" level is reached), the following reactor coolant system response would result:

Maximum Thot -- 6020F Maximum RC pressure -- 2415 psig (Does not include initial pressure peak of 2500 at time of raactor trip which is unaffected by EFW flow) j Maximum pressurizar level -- 255 inches No actuation of safety valves, although the PORV cycles twice.

We would consider this system response acceptable.

B&W document #86-1152597-00 contains the results of an analysis performed for SMUD which has a core power of 2772 MWt the same as Davis-Besse. The objective of this analysis was also to predict RC system response with an AFW flow rete of 560 gym. The analytical assumptions of interest for this case are as follows:

o Initial power -- 102%, 2827 MWt o Decay heat (1979 ANS 5.1 decay heat) x 1.2 h ..

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o RC pumps running o Reactor trip on high pressure o Arw delivery to steam generator 70 seconds after reactor trip

For this case the pressurizer safety valves were predicted to actuate, 8.owever the system overheating was terminated prior to the pressurizer going solid and passing water through the valves.

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