ML20062F539
| ML20062F539 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/11/1978 |
| From: | Roe L TOLEDO EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| FOIA-79-98, TASK-TF, TASK-TMR TAC-11050, NUDOCS 7812190081 | |
| Download: ML20062F539 (3) | |
Text
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. Docket No. 50-346 t
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License No. NPF-3 i
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Serial No.'471 December 11, 1978 Director of Nuclear Reactor Regulation Attention:
Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors U. S. Nuclear Regulatory Cornission Washington, D. C. 20553
Dear Mr. Reid:
In response to the request by your Mr. Guy Vissing in his telephone call on December 8, 1978, attached is our analysis scoporting continued operation of Davis-Besse Nuclear Power Station Unit I with dual level setpoint co'ntrol of the steam generators.
1 Yours very truly,
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THE TCLECO ECISCN COMPANY EDISCN PLAZA 3C0 MACISCN AVENUE TCLECO, CH;O 43652
Dockst No. 50-346
- License No. NPF-3 Serial No. 471 December 11, 1978 i
Analysis Supporting C<ntinued Operation of Davis-Besse Nuclear Power Station Unit I with Dual Level Setpoint Control of the Steam Generators The Davis-Besse Unit 1 procedures, as recently revised, require manual control of steam generator (SG) level at 35 inches on the startup range level indicators, following anticipated operational cecurrences where auxiliary feedwater is required to remove decay heat. This SG level was chosen to provide adequate natural circu-lation capability and additional margin for the maintenance of indicated pressurizer level.
The results of the natural circulation test at Davis-Besse Unit 1 (TP 800.04) show that there is 4.6% of full reactor coolant system-(RCS) flow with the start-up range level indicators indicating 35" of water in each of the steam generators. The minimum acceptable flow is 1.63% at the test power level of above 3.85% of full power.
Therefore, under any condition except when a small RCS break occurs,the station can safely operate with an indicated SG 1evel of 35" on the start-up range level indica-tors. This is a change from SG levels as described in the Davis-Besse Unit 1 FSAR, and this safety evaluation is considered the 10CFR50.59 review of this change.
Manual control of the SG level at 35" indicated is a temporary measure until the
" auto-essential" level control is modified with a dual setpoint. Since November, 1977 the Davis-Besse Unit 1 operators have successfully manually controlled the steam generator levels on each occasien that auxiliary feedwater was automatically started by the steam and feedwater rupture control system (SFRCS).
f Manual control does not reduce the capability of each auxiliary feedwater pump (AFP) to deliver 800 gpm to each SG.
At a 35" indicated level an average total flow capa-city of less than 400 gpm per SG with both SGs in service is required to remove the decay heat from the RCS.
The " auto-essential" level control was included in the design of Davis-Besse Unit 1 in 1976 to prevent a SG from flooding if two AFPs were feeding it.
Prior to this time the design provided for automatic SG 1evel control only from the integrated i
control system (ICS). The only essential level control was manual and when SFRCS initiated auxiliary feedwater flow, the AFP turbine was started at full speed with the SG level control in the essential manual mode. In 1976, it was determined
- that when two AFPs were running at full speed and supplying a single SG,' that SG would flood in less than ten minutes if an operator failed to take manual action to control the SG level.
(Two AFPs will supply one SG only after a steam line break).
With the " auto-essential" level control a single active failure of one level controller (which allows one AFP to continue running at full speed with the SG 1evel above 120"), the SG will become flooded in approximately 15 minutes. This allows ample time for operator intervention.
I v
Docket No. 50-346 License No. NPF-3 Serial No. 471 December 11, 1978 Page Two The B&W accident analysis for a small RCS break at Davis-Besse Unit 1 (3AW-10075A, Revision 1. March, 1976) assumes that high pressure injection has been started by the safety features actuation system, a loss of off-site power has occurred, and the SGs have been supplied with auxiliary feedwater. B&W advises that this analysis is applicable for nominal SG level as low as 120".
Because of the present time restraints anticipated for additional analyses of small RCS breaks with SG level lower than 120", it has been decided to provide dual level setpoint control of the SGs.
There has been an analysis of the pressurizer levels during various reactor trips, including the November 11, 1977 incident during which all frur reactor coolant pumps were tripped with the reactcr operating about 40% of full power. This analysis indicates that the pressurizer level reached 29" below the low level sensing tap (the sensing tap is 117" above the horizontal centerline of the 10" pressurizer surge line).
The analysis shows that for a reactor trip from 15% of full power with the loss of all reactor coolant pumps and main feedwater pumps (as will occur in the loss of off-site power test), pressurizer level indication could be lost if about 100 inches of auxiliary feedwater level are added to each SG and also causing the steam pressure to decrease to 770 psig.
In contrast, maintaining a 35" SG 1evel results in a pressurizer indicated level of approximately 84" and a steam pressure of 950 psig.*
Considering that a 120" SG level is only required for a small RCS break, and the desire to maintain pressurizer level indication during transients in acecrdance with General Design Criterion 13, it has been determined that dual level setpoints are required.
When the modification is completed to provide the dual setpoints for the " auto-essential" level control on the SGs, all manual control of the SC levels when on auxiliary feedwater will be eliminated from the Davis-Besse Unit 1 procedures.
This change is not an "Unreviewed Safety Question".