ML20078L952
| ML20078L952 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 11/29/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078L955 | List: |
| References | |
| NUDOCS 9412010256 | |
| Download: ML20078L952 (3) | |
Text
I I
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
a.
Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and b.
One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.*
f APPLICABILITY: MODES 1, 2, and 3.
ACTION:
a.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hour: 'nd in HOT SHUTDOWN within the following 6 i
hours.
c.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to operable status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
a.
At least once per 31 days by:
1)
Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; 2)
Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER; and 3)
Verifying that the isolation valves in the auxiliary feedwater suction line from the upper surge tanks are open with power to the valve operators removed.
- Not applicable with steam pressure less than 900 psig.
McGUIRE - UNITS 1 and 2 3/4 7-4 Amendment No.151 (Unit 1)
Amendment No. 133 (Unit 2) 9412010256 941129 PDR ADOCK 05000369 p
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
At least once per 92 days on a STAGGERED BASIS by:
1)
Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1210 psig at a flow of greater than or equal to 450 gpm; and 2)
Verifying that the steam turbine-driven pump develops a dis-charge pressure of greater than or equal to 1210 psig at a flow of greater than or equal to 900 gpm when,the secondary steam supply pressure is greater than 900 psig.
The provisions of Specification 4.0.4 are not applicable for entry 11to MODE 3.
c.
At least once per 18 months during shutdown by:
1)
Verifying that each automatic valve in the flow pat h actuates to its correct position upon receipt of an Auxiliary feedwater Actuation test signal, 2)
Verifying that each auxiliary feedwater pump starts a> designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal, and 3)
Verifying that the valve in the suction line of each auxiliary feedwater pump from the Nuclear Service Water System automati-cally actuates to its full open position within less than or equal to 13 seconds on a Low Suction Pressure test signal.
This verification is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving greater than or equal to 900 psig in the secondary side of the steam generator.
J McGLIRE - UNITS 1 and 2 3/4 7-5 Amendment No. 151 (Unit 1)
Amendment No. 133 (Unit 2)
PLANT SYSTEMS BASES SAFETY VALVES (Continued) 109
- Power Range Neutron Flux-High Trip Setpoint for four loop operation,
- Maximum percent of RATED THERMAL POWER permissible by P-8 Setpoint for three loop operation.
This value left blank pending NRC approval of three loop operation, X - Total relieving capacity of all safety valves per steam line in lbs/ hour, and Y - Maximum relieving capacity of any one safety valve in lbs/ hour.
3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating' conditions in the event of a total loss-of-offsite power.
Each electric motor-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1210 psig to the entrance of the steam generators.
The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 900 gpm at a pressure of 1210 psig to the entrance of the steam generators. This capacity is sufficient-i to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the RHR System may be placed into operation.
Verification of the steam turbine-driven pump discharge pressure should be deferred until suitable test conditions are established (i.e., greater than or equal to 900 psig in the secondary side of the steam generator).
This deferral is required because until 900 psig is reached, there is insufficient steam pressure to perform-the test.
3/4.7.1.3 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.
This dose also includes the effects of a coincident 1.0 gpm reactor-to second-ary tube leak in the steam generator of the affected steam line.
These' values are consistent with the assumptions used in the accident analyses.
1 5
McGuire - UNITS 1 and 2 B 3/4 7-2 Amendment No.
151 (Unit 1)
Amendment No.
133 (Unit 2)
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