ML19320C971
| ML19320C971 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/20/1980 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| IEB-80-06, IEB-80-6, LAC-6987, NUDOCS 8007180362 | |
| Download: ML19320C971 (5) | |
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s D DA/RYLAND
[k COOPERAT/VE. eo sox 8i7 2615 EAST AV SOUTH. L A CROSSE WISCONSIN 54601 (608) 788 4 000 June 20, 1980 In reply, please refer to LAC-6987 DOCKET NO. 50-409 Mr. James G.
Keppler Regional Directer U.
S.
Nuclear Regulatory Commission Directorate of Regulatory Operations Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 IE BULLETIN NO. 80 ENGINEERED SAFETY FEATURE (ESP) RESET CONTROLS
Reference:
(1)
NRC Letter, Keppler to Linder, dated March 13, 1980.
Dear Mr. Keppler:
Your request (Reference 1) for a review of engineered safety feature reset control has been completed.
All drawings for systems serving safety related functions were re ~
viewed at the schematic level to determine whether or not upon the reset of a safety actuation signal, all associated safety related equipment remains in its emergency mode.
Except for the equipment listed in Appendix A of this letter, no safety-related functions were noted~to reset upon the removal of the actuating signal.
Tests were performed during the recent outage to demonstrate whether or not equipment remains in its emergency mode upon removal of the actuating signal.
Functions included in this test were containment isolation, onsite AC emergency diesel generators and emergency core cooling systems.
The equipment listed in Appendix A, an attachment to this letter, does not by design remain in its emergency mode.
Proposed modifica-tions where they are desirable are defined. 8007180360dk UUN 2 3 1980 Q
Mr. James G. Keppler LAC-6987 Regional Director June 20, 1980 Approval for this response to' be submitted beyond the due date was granted by Mr. Ken Baker on June 16, 1980.
If there'are any questions, please. contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE
)
~ A r.v A
n Frank Linder, General Manager FL:JDP:af Attachment cc:
U.
S.
Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D.
C.
20555 STATE OF WISCONSIN )
)
COUNTY OF LA CROSSE)
Personally came before me this d#
day of June, 1980, the above named Frank Linder, to me known to be the person who executed the foregoing instrument and acknowledged the same.
? h)L< b w
Wisconsin.gyic, Notary Pul La Crosse County My Commission Expires 2/26/84.
i APPENDIX A 1.
Low Pressure Emergency Core Spray Valve No. 53-25-001.
A low reactor water level signal (-12 inches) simultaneous with reactor pressure not more than 30 psig opens this valve to actuate a low pressure injection by gravity.
The valve has a hand controller in the control room.
In case of loss of nitrogen or loss of 120V AC non-interruptible control power to the solenoid controlling nitrogen to operate the valve, it will fail open.
The removal of either low reactor water level or reactor pressure signal reverses the valve opening action.
This is the desired mode of operation of this valve and no modification is proposed.
2.
Alternate Core Spray Valves No. 38-30-001 and No. 38-30-002.
Two parallel control valves (No. 38-30-001 and No. 38-30-002) are provided in the line to the reactor.
These valves both open automatically when the reactor water level drops below the low level scram set point (-12 inches) and containment building pressure is equal to or greater than 5 psig.
The water supply for the alternate core spray system is furnished by two diesel pumps which start upon receiving a signal of high containment building pressure (5 psig). Alternate core spray flow to the vessel commences when reactor vessel pressure drops to approximately 150 psig.
Removal of the low water level signal closes these valves by I
design.
This is the desired mode of operation and no modifi-cations are necossary.
3.
Main Steam Isolation Valve Bypass Valve No. 64-24-030.
The bypass valve is a diaphragm-operated flow control valve which is controlled from the control room.
The valve is closed automatically on low reactor water level, low main condenser vacuum or low steam pressure at the turbine inlet.
When any one of the above signals is removed and the valve position is manually controlled to a specified opening, the valve will reopen.
There appears be no necessity for the bypass valve to auto-matically reopen, therefore the control features for this valve will be modified at the next refueling outage to permit the l
valve to remain in its emergency mode upon removal of the l
actuating signal.
l 4.
Shutdown Condenser.
The shutdown condenser operates automatically upon high primary system pressure (1325 psig) on either one of two safety channels, or when either the main steam isolation valve or the turbine building isolation valve leaves the full open position.
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When the shutdown condenser operation is initiated by a high primary system pressure, the system functions as a pressure control starting when reactor pressure reaches 1324 psig and stopping when pressure is reduced to 1300 psig.
A start signal opens:
Parallel Steam Inlet Valves (62-25-001 and 62-25-011),
Parallel Condensate Valves (62-25-002 and 62-25-012),
Off-Gas Vent Valve (62-25-003) which remains open for two minutes to vent non-condensibles and then closes; and closes:
Drain Trap Isolation Valve (62-25-017).
The water supply to the shell side of the condenser cycles auto-matically to retain a specified level during operation.
The principle water supply make-up is demineralized water (Valve No.
62-25-004) and the automatic back supply is high pressure service water (Valve No. 62-25-005).
This pressure control mode of operation is by design and no modifications are required.
When the shutdown condenser operates due to closure of either the main steam isolation valve or the turbine building isolation valve.
Manual action is required to terminate operation.
5.
Screen Wash Control.
The main circulating water system has traveling screens which remove debris from the main condenser cooling water.
These smreens are periodically washed by high pressure service water via two in parallel valves (:No. 75-25-021 and 75-25-022).
These valves are closed if high reactor building containment pressure
( +r psig) occurs to reduce demand on the alternate core spray system. When the high reactor pressure signal is removed, the valves return to their automatic cycling position (screen wash is in progress 10 minutes out of each hour).
A review of this design feature reveals no overriding need to have the screen wash valves return to automatic cycling, therefore, this feature will be modified at the next refueling outage.
6.
Low Voltage Relay lA and 1B 440 Volt Essential Buses.
The undervoltage relays are 440 volt essential bus lA (device numbers 427ESA Phase A and 427ESA Phase C) and 440 volt essential bus 1B (device numbers 427ESB Phase A and 427ESB Phase C) reset once the low voltage condition no longer exists.
An undervoltage condition on either bus initiates a reactor scram.
An undervoltage condition on the 1A bus starts the lA Emergency Diesel Generator.
An undervoltage condition on the 1B bus starts the 1B Emergency Diesel Generator.
The reset of the relays is by design and does not reverse any of these actions started by the low voltage signal.
No modifi-cations are required.
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