ML19225A335
| ML19225A335 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/21/1979 |
| From: | Walbridge W SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 7907180871 | |
| Download: ML19225A335 (5) | |
Text
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SACRAM2NTO Ml'NICIPAL UTILITY DISTRICT E] 6201 S streat, Box 15330, sacramento, California 95 May 21, 1979 Mr. R. H. Engelken, Director Region V Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission 1990 North California Boulevard
'hlnut Creek Plaza, Suite 202 Walnut Creek, California 94596 Docket No. 50-312 Rancho Seco Nuclear Generating Station, Unit No. 1
Dear Mr. Engelken:
The Sacramento Municipal Utility District has responded to items 1, 2, 3, 4, and 6 of IE Bulletin No.79-05B dated April 21, 1979 in letters dated April 22, 1979 and May 2, 1979.
This letter provides the District's responses to the remaining items, numbers 5 and 7.
Item 5 Provide for NRC approval ? design review and schedule for implementation of a safety gradt automatic anticipatory reactor SCRM for loss of feedwater, turbine trip, or significant reduction in steam generator levels.
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Response to Item 5 An evaluation for anticipatory trips on turbine trip, loss of main feedwater, and low steam generator level has been completed.
This evaluation showed a low steam generator level not to be anticipatory and, therefore, it has not been included as an anticipatory trip function.
A review of site data and calculations by the Babcock & Wilcox Company shows that a reactor trip on high reactor coolnnt pressure would precede a steam generator level trip for a level setpoint that would not interfere with normal operations and maneuvers.
However, anticipatory trips for loss of feedwater and turbine trip caq be designed to trip the reactor in a more expedient manner than the high reactor coolant pressure trip for some overheating transients.
Such an antici-patory trip will provide more margin t'o PORV setpoint during the initial overpressurization resulting from loss of feedwater and/or turbine trip.
These trips wiil provide slightly more time to POVP setpoint and pressurizer fill for delayed auxiliary feedwater initiation conditions.
409 343 7907180 2 7f
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A fi LLECTalt SYSTI" 3 E l' V l f G PORI THfN 600,001 I f, THi H L A i: I L i F O R fi i A
i f1r. R. H. Engel ken May 21, 1979 Figures 1 and 2, attached to this letter. provide simplified drawings of the reactor trip on loss of main feedwater and turbine trip respectively. Control oil pressure switches on both main feedwater pumps will input an open indication to the 7PS on feed-water pump trip.
Contact buffers in the R?S will sense the cont a inputs and initiate an RPS trip when both pumps have tripped.
This trip will be bypassed below a predetermined flux level, typically 20% full power.
Contact outputs fram the main turbine electro-hydraulic control unit will input an open indication to the RPS on turbine trip.
Contact buf fers in the RPS will sense the contact inputs and initiate an RPS trip when a turbine trip is indicated.
This trip will also be bypassed below a predetermined flux level, typically 70% full power.
A more detailed design and safety evaluatica will be performed as required by our long term commitments in the liRC Order of May 7, 1979.
It is anticipated that at least nine months will be required to procure the necessary equipment following fiRC approval of the design. The District proposes to install such a modification during the first refueling outage following the receipt of the required equipment.
Item 7 Propose changes as required to those technical specifications which must be modified as _ result of your implementing the above items.
Respor.se to Item 7 The District feels that technical specification number 2.3.1 is the only one which could possibly be modified as a result of implementing changes required by IE Bulletin 79-05B.
This sp ;ification providas for a high pressure reactor trip at a maximum of 2355 psig.
Item 3 of IE Bulletin 79-05B resulted in a decrease to that cetpoint to 2300 psig.
The District does not feel it is proper to change this te~chnical spacification because thr present specification allows for this lowered setpoint.
In ad;ition the hard wire control grade trips of the reactor upon a turbine trip or loss of feedwater, as provided by the District in response to the flRC Order of May 7,1979, and the safety grade reactor trip to be added for the same events, provide the prcmpt termination of nuclear heat generation for these transients.
The lowered 2300 psig reactor pressure trip point is not a true Limiting Safety System Setting providing a required margin of safety to a Safety Limit as does the present technical specification 409 E9
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Mr. R. H. Engelken ilay 21, 1979 limit of 2355 psig.
Instead, this lowered setpoint provides increased margin to prevent operation of the PORV.
For that reason the District does not propose any changes to the technical specifications at this time.
We do feel that it will be appropriate to add tec".i.ical specifications for the reactor trip on loss of main feedwater or turbine trip when these safety grade modifications are made to the plant.
We feel that this letter provides the remainder of the responses required by IE Bulletin 79-95B.
However, if we can provide any additional information, please advise.
Sincerely yours,.
'll I )L /I..e Si', 'I c
Wm. '. Walbridge g
General f13 nager c: John G. Davis Acting Director, Division of Reactor Operations Inspections 409 3:0
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