ML19270F332
| ML19270F332 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/30/1979 |
| From: | Delgeorge L COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 7902060267 | |
| Download: ML19270F332 (2) | |
Text
Commonwealth Edison One First National Plaza Chicago, !!!inois Address Reply to: Post Office Box 767
. Chicago, Illinois 60690 January 30, 1979 Mr. James G. Kappler,-Director Directorate of. Inspection and Enforcement - Region III U.S. Nuclear Regulatory Ceaunission.
799 Roosevelt Road Glen Ellyn, Illinois' 60137
Subject:
LaSalle County Station Units 1 and 2
. Safety Relief Valve Control System NRC Docket Nos. 50-373 and 50-374 References (a):
M. S. Turbak-letter to J. G. Kappler dated November 11, 1977 (b) :
C. Reed letter to 0. D. Parr dated July 18, 1978 Dear Mr. Keppler On October 14, 1977,- Caussonwealth Edison notified the NRC of a possible safety / relief valve control deficiency at LaSalle County Station under 10 CFR 50.55(e).
This reportable condition was basea on the transient analysia' prediction.of the sequence and number of safety relief valves expected to operate following a reactor isolation event.
A preliminary report on this subject was subr itted in Reference (a).
To investigate the effects of subsequ1nt actuation,a plant unique analysis was performed for the LaSalle County Station and the results presented in-the LaSalle County Station Design Assessment Closure Report, Section'3.2.2, Reference (b).
The report stated that only six safety relief valves are subsequently actuated (second pop).when the pressure differential between opening and closing setpoint}is 10,0 psi.. This in only one-third of the total number.of ~ aafety relief valves '(18) which was used as a design basis.
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,'s NRC Docket Nos. 50-373/374 Mr. James G. Kappler:
2-January 30,,1979 With the-incorporation of the safety relief T-quencher device desi~ned by IOfU it is anticipated that the resulting loads, g
due to subsequent SRV actuation, will be mitigated and that the actual load magnitude for subsequent SRV actuation will be.less
' than the current design basis.
In order to insure-this result, the SRV closing setpoints will be~ adjusted to provide for a 100 psi blowdown in order to confona to the analysis, results.
This modification will be implemented prior to fuel loading on Lasalle Units 1.and.2
. No further modification to the SRV control system is necessary.-
Please address any. questions on this subject to this office._ As has been stated, this represents a final-reporti in accordance with the requireements of 10 CFR 50.55(e).
Very truly yours, L. O. De1 George Nuclear Licensine Administrator Boiling water Reactors cc C. Reed i
Director Inspection and Enforcement a
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