ML20003F491
| ML20003F491 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/16/1981 |
| From: | Delgeorge L COMMONWEALTH EDISON CO. |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8104210390 | |
| Download: ML20003F491 (3) | |
Text
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Commonwealth Edison One First Nat:oeat Piata. Chicago. Illmois Aadress Reply to: Post Off:ce Box 767 Chicago, Ilhnois 60690 N) r ll h D%
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Apri1 16, 1981
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Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
LaSalle County Station Unit 1 Initial Test Program - Special Test Simulated Loss of AC Power NRC Docket No. 50-373 l
Reference (1):
L. O. DelGeorge letter to B. J. Youngblood (LOD 81-40-05); dated February 9, 1981.
Dear Mr. Youngblood:
As was indicated in Ref erence (1) Commonwealth Edison has committed to perform a special test simulating station blackout (loss of all AC power) conditioned on the results of a detailed safety-evaluation concurred in by Commonwealth Edison and General Electric and accepted by the NRC under the requirements of 10 CFR 50.59.
Enclosed for your review is a detailed outline for that
" blackout" test, which discusses the methods to be used, equipment assumed to be operable and conditions necessary to terminate testing.
In order for the detailed test procedure and safety evaluation to be developed this test outline must be finalized.
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Therefore, any questions or comments you may have on this outlin3 are requested by May 15, 1981.
Because of the constaints assoicated with scheduling General Electric engineering support, comments received from you after May 15, may severely effect the schedule for performante of this test, Very truly yours, L. O. DelGeorge Nuclear Licensi;.g Administrator tnclosure pl S
cc:
NRC Resident Inspector - LSCS 2376B II sla4210 % C
ELACKOUT TEST PURPOSE To monitor the reactor and containment response to a simulated loss of site AC power.
METHOD After greater than one week of continuous high power operation (180%), the reactor will be scram =ed and containment isolation initiated. The following AC powered systems will be secured or placed in Pull-to-Lock (as appropriate) to prevent them from affecting plant response during the transient:
1.
Feedwater 2.
RHR (all mades) 3.
HPCS 4.
LPCS 5.
Primary Containment Ventilation 6.
Primary Containment Chilled Water 7.
IN Compressors 8.
SA/IA Backup Supply to IN 9.
MSIV's 10.
Reactor Recirculation Pumps Isolating the above systems will allow only the DC powered systems to affect the reactor and containment response, except for the CRD system.
Continued operation of the CRD system during this test is necessary to prevent damage or degradation of the Control Rod seals and Recirculation Pump seals. The relatively small amount of water injected by the CRD pumps is not ccasidered to be enough to significantly influence the test results.
ANTICIPATED RESULTS Reactor Pressure and te=perature rapidly increase, causing most or all of the SRV's to lift. The Low Level System (LLS) SRV logic,will engage, and the SRV's will reduce pressure to 896 psig.
The SRV's will then maintain reactor pressure between 896 and 1006 psig. RCIC, which shall be operating in a test mode prior to the scram, will be manually lined up to the vessel.
Initially RCIC flow will be inadequate to make up for SRV flow rate, and vessel water level will decrease.
After about 5 minutes, with water level around -120 inches (four feet over the core),
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RCIC flow will exceed the SRV average flow and level will slowly start to recover.
Level should be restored to normal about one hour into the test.
During this test, the suppression pool' water temperature will increase gradually due to the addition of steam from the SRV's and RCIC exhaust.
This increase will be on the order of 300 F in the first hour and fifteen minutes.
Due to the loss of drywell cooling, drywell pressure and temperature will increase throughout this test.
Blackout Test Page 2 TEST TERMINATION The test will be terminated and plant conditions returned to nor=al upon reaching any of the following conditiens:
.e inside the containment (5 gpm increase).
1.
Any observed increase of leak 2.
Reactor water level decreases te -129".
3.
Suppression pool temperature res..aes 110 F.
4.
Drywell temperature reaches 200 F.
5.
Containment pressure reaches 5 psig.
6.
RCIC trips and cannot be restarted within 1 minutes.
7.
Any SRV sticks open.
RETURN TO NORMAL Upon termination of the test, the following actions (as appropriate) will be taken to return the plant to a normal hot shutdown condition:
1.
Start the MDRFP and restore vessel level to normal. Avoid exceeding 100 F temperature change in any one hour period.
2.
If the MDRFP cannot be started, use HPCS to refill.
3.
Initiate suppression pool cooling with both RHR loops.
4.
Restart containment ventilation fans.
5.
Restore PCCW flow.
The jumpers required to override the 1.69 psig D/W pressure isolation should be installed prior to initiating the test.
ACCEPTANCE CRITERIA There are no acceptance criteria asscciated with this tes*;.
Plant response to this transient will be evaluated and disseminated.
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