ML17326A676
| ML17326A676 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/08/1980 |
| From: | Dolan J INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8004280398 | |
| Download: ML17326A676 (6) | |
Text
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INDIANA L MICHIGAN ELECTRIC COMP NY P. 0.
BOX 18 Bo WLING G R E EN ST ATION NEW YORK, N. Y. 10004 April S, 19SO AEP: NRC: 00291A Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2
Docket Nos.
50-315 and 50-316 Licerjse Nos.
Subject:
IE Bulletin 80-04:
Analysis of,a PWR Main Steam Break with Continued Feedwater A'ddition Mr. James G. Keppler, Director U.S. Nuclear Regulatory Commission, Region III 799 Roosevelt Road Glen Ellyn, Illinois'0137
Dear Mr. Keppler:
This letter responds to IE Bulletin No. 80-04 received on February 13, 1980.
The Bulletin addressed possible non-con-servative assumptions in the calculation of reactor system and containment responses to steam line break events.
The attach-ment to this letter summarizes the review which was performed for the Donald C. Cook Nuclear Plant.
On the basis of the review; we have concluded that our present steam line break containment and reactor system. analyses as reported in the Cook Plant FSAR adequately address the concerns of'he bulletin.
Very truly yours, JED:em ohn E. Dolan Vice President N
cc:
R.
C. Callen G. Charnoff R. S. Hunter R.
W. Jurgensen D. V. Shaller - Bridgman Office of Inspection and Enforcement NRC
l~
i
EVALUATION OF STEAM LINE BREAK EVENTS FOR DONALD C.
COOK PLANT FOR IE BULLETIN 80-04 w
The containment and Reactor Coolant System (RCS) responses to postulated Steam Ltne Break"(SLB) Events have'been analyzed in considerable detail for the Donald C.
Cook Nuclear Plant.
Re-sults of these analysis are reported in FSAR sections 14.2.5 (both Units 1
and 2) and Appendix g (questions 22.9, 212.7, 212.23, 212.24, 212.25, 212.34, 212.35 and 212.39).
Further information has been supplied-in our transmittal No. AEP:NRC:00131 dated April 1, 1980.
A review of the containment response analysis (in particular 922.9) reveals that the potential impact of various flow control failures have been addressed for the Cook Plant.
A series. of break sizes and initial-reactor power levels were analyzed in conjunction with potential failures such as main steam isolation valve (MSIV) failure, feedwater isolation valve failure, failure of main feedwater pump to trip, or failure of the auxiliary feedwater runout control system, in order to determine the most severe break cond1tions for the containment pressure and temperature response.
The analysis 1ndicated that the worst large SLB 1s a
1.4 ff break occuring at 102% power in conjunction with a NSIV.
failure.
The worst spl1t break (small break) is a 0.942 ft2 severance pccuring at 305 power in conjunction with the failure of the auxiliary feedwater runout control system.
In all cases
- analyzed, the combined mitigating'ctions of the ice condenser, containment spray system and the passive structural heat sinks con-trolled the transient such that a containment overpressure condition was never calculated to occur.
Item 2 of the Bulletin addresses the core power tpansieqts associated with SLB events.
Westinghouse has reviewed the assumptions made for main and auxiliary feedwater flow during the event.
The following input assumptions made in the Cook Plant analysis are relevant to the current concerns.
1.
The reactor is assumed to be initially in a hot shutdown con-dition at the minimum allowable shutdown margin.
Further, re-activity response parameters were chosen as to both minimize the moderator temperature coefficient and underpredict beneficial feedback effects such as Doppler and void coefficients.
Minimum boron injection capability concurrent with the most restrictive single failure in the safety injection system was assumed.
2.
In the analysis of a double ended SLB,allmain feedwater flow is assumed from the beginning of the transient at a very conservative cold temperature thereby maximizing the RCS cooldown.
h
3.
All auxiliary feedwater pumps -are ini'tially assumed to be operating in additi'on to the main feedwater pumps.
The initial flow is equivalent to the rated flow of all pumps at the steam generator design pressure.
4.
Feedwater is assumed to continue at its initial flow rate until feedwater isolation is complete approximately ten seconds after, the break occurs, while auxiliary feedwater is assumed to continue at its initial flow rate. Isolation of main feedwater is redundant in the Cook Plant design in that in addition to the normal control actions, a safety injection signal will rapidly close all feedwater control valves, backup feedwater isolation valves, trip the main feedwater pumps and close the main feedwater pump discharge valves.
5.
Main feedwater is completely terminated following feedwater isolation.
The analyse's'erformed for the Cook Plant show the core transient results to be insensitive to auxiliary feedwater flow assumptions.
The first minute of the transient is dominated entirely by the steam flow contribution to primary-secondary heat transfer which is the forcing function for both the reactivity and thermal hydraulic transients.
The effect of the auxiliary feed-water runout control system is minimal.
Greater feedwater flows accelerate automatic safeguards actuation (i.e
, steamline and feedwater isolation, safety injection, etc.).
Therefore, the assumptions listed above are both appropriate'nd conservative for the short term aspect of the SLB transient.
The auxili'ary feedwater flow becomes a dominant factor in de-termining the duration and magnitude of the steam flow transient duri.ng the later stages of the event.
However, the limtti'ng core conditions occur duri'ng the fi'rst minute of the transient due to the higher steam flows inherently present ear ly in the event and the introduction of boron into the core.
In the worst scenario analyzed for the Donald C.
Cook Plant, whtch was a complete severance of the steam line at the steam generator outlet with offsite power available, the peak core power never. exceeded 255 of full power and DNB was not calculated to occur, In conclusion, revi'ew of the relevant analyses already performed and reported for the Cook Plant for the steam line break provide adequate assurance that the licensing basis remains valid in light of the concerns of IE Bulleti'n 80-04.
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