ML20204J375
| ML20204J375 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/31/1986 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20204J364 | List: |
| References | |
| NUDOCS 8608110074 | |
| Download: ML20204J375 (4) | |
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1 PROPOSED CHANGES TO ANO-1 TECHNICAL SPECIFICATIONS i
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- 3. 5.1. 7 The Decay Heat Removal System isolation valve closure setpoints shall be equal to or less than 340 psig for one valve and equal to or less than 400 psig for the second valve in tha suction line.
The relief valve setting for the DHR system shall be equal to or less than 450 psig.
3.5.1.8 The degraded voltage monitoring relay settings shall be as follows:
a.
The 4.16 KV emergency bus undervoltage relay setpoints shall be >3115 VAC but (3177 VAC, b.
The 460 V emergency bus undervoltage relay setpoints shall be
> 423 VAC but (431 VAC with a time delay setpoint of 8 seconds 1 second.
3.5.1.9 The following Reactor Trip circuitry shall be operable as indicated:
1.
Reactor trip upon loss of Main Feedwater shall be operable (as determined by Specification 4.1.a and item 35 of Table 4.1-1) at greater than 5% reactor power.
(May be bypassed up to 10% reactor power.)
2.
Reactor trip upon Turbine Trip shall be operable (as i
dettermined by Specification 4.1.a and item 41 of Table 4.1-1)
I at greater than S% reactor power.
(May be bypassed up to 45%
l reactor power.)
3.
If the requirements of Specifications 3.5.1.9.1 or 3.5.1.9.2 cannot be met, restore the inoperable trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or bring the plant to a hot shutdown condition.
3.5.1.10 The control room ventilation chlorine detection system instrumentation shall be operable and capable of actuating control room isolation and filtration systems, with alarm / trip setpoints adjusted to actuate at a chlorine concentration of $5 ppm.
3.5.1.11 For on-line testing of the Emergency Feedwater Initiation and Control (EFIC) system channels during power operation only one channel shall be locked into " maintenance bypass" at any one time.
If one channel of the NI/RPS is in maintenance bypass, only the corresponding channel of EFIC may be bypassed.
3.5.1.12 The Containment High Range Radiation Monitoring instrumentation shall be operable with a minimum measurement range from 1 to 107 R/hr.
Amendment No. 60, 61, 69, 91, 94 42a r-p m
y
for protective action from a digital ESAS subsystem will not cause that subsystem to trip.
The fact that a module has been removed will be continuously annunciated to the operator.
The redundant digital subsystem is still sufficient to indicate complete ESAS action.
The testing schemes of the RPS, the ESAS, and the EFIC enables complete system testing while the reactor is operating.
Each channel is capable of being tested independently so that operation of individual channels may be evaluated.
The EFIC is designed to allow testing during power operation.
One channel may be placed in key locked " maintenance bypass" prior to testing.
This will bypass only one channel of EFW initiate logic.
An interlock feature prevents bypassing more than one channel at a time.
In addition, since the EFIC receives signals from the NI/RPS, the maintenance bypass from the NI/RPS is interlocked with the EFIC.
If one channel of the NI/RPS is in mainenance bypass, only the corresponding channel of EFIC may be bypassed.
The EFIC can be tested from its input terminals to the actuated device controllers.
A test of the EFIC trip logic will actuate one of two relays in the controllers.
Activation of both relays is required in order to cctuate the controllers.
The two relays are tested individually to prevent automatic actuation of the component.
The EFIC trip logic is two (one-out of-two).
Reactor trips on loss of all main feedwater and on turbine trips will sense the start of a loss of OTSG heat sink and actuate earlier than other trip signals. This early actuation will provide a lower peak RC pressure during the initial over pressurization following a loss of feedwater or turbine trip event. The LOFW trip may be bypassed up to 10% to allow sufficient margin for bringing the MFW pumps into use at approximately 7%.
The Turbine Trip trip may be bypassed up to 45% based on BAW-1893, " Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip," October 1985 and the NRC Safety Evaluation Report for BAW-1893 issued from Mr. D. M.
Crutchfield to Mr. J. H. Taylor via letter dated April 25, 1986.
The Automatic Closure and Isolation Eystem (ACI) is designed to close the Decay Heat Removal System (DHRS) return line isolation valves when the Reactor Coolant System (RCS) pressure exceeds a selected fraction of the DHRS design prusure or when core flooding system isolation valves are opened.
The ACI is designed to permit manual operation of the DHRS return line isolation valves when permissive conditions exist.
In addition, the ACI is designed to disallow manual operation of the valves when permissive conditions do not exist.
Power is normally supplied to the control rod drive mechanisms from two separate parallel 480 volt sources.
Redundant trip devices are employed in each of these sources.
If any one of these trip devices fails in the untripped state, on-line repairs to the failed device, when practical, will be made and the remaining trip devices will be tested.
Four hours is ample time to test the remaining trip devices and, in many cases, make on-line repairs.
Amendment No. 50, 60, 61, 91 43a i
TABLE 3.5.1-1 (Cont'd) 12.
With the number of operable channels less than required, either return the indicator to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or verify the block valve closed and power removed within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If i
the block valve cannot be verified closed within the additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, de-energize the electromatic relief valve power supply within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
E.
13.
Channels may be bypassed for not greater than 30 seconds during reactor coolant pump starts.
If the 5
automatic bypass circuit or its alarm circuit is inoperable, the undervoltage protection shall be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, otherwise, Note 14 applies.
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a 14.
With the number of channels less than required, restore the inoperable channels 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 hot shutdown 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 cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
15.
This trip function may be bypassed at up to 10% reactor power.
16.
This trip function may be bypassed at up to 45% reactor power.
17.
With no channel operable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the inoperable channels to operable status, or initiate
~F and maintain operation of the control room emergency ventilation system in the recirculation mode of 9'
operation.
ld 18.
With one channel inoperable, restore the inoperable channel to operable status within 7 days or within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.
19.
This function may be bypassed below 750 psig OTSG pressure.
Bypass is automatically removed when pressure exceeds 750 psig.
20.
With one channel inoperable, (1) either restore the inoperable channel to operable status within 7 days, or (2) prepare and submit a Special Report to the Commission pursuant to Specification 6.12.4 within 30 days following the event, outlining the action taken, the cuase of the inoperability, and the plans and schedule for restoring the system to operable status.
With both channels inoperable, initiate alternate methods of mor'toring the containment radiation level within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in addition to the actions described above.
21.
With one channel inoperable, restore the inoperable channel to operable staus within 30 days or be in hot shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless containment entry is required.
If containment entry is required, the inoperable channel must be restored by the next refueling cutage.
If both channels are inoperable, restore the inoperable channels within 30 days or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.