ML19262C679
| ML19262C679 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/08/1979 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19262C672 | List: |
| References | |
| NUDOCS 8002150386 | |
| Download: ML19262C679 (7) | |
Text
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3.4 STEAM AND POWER CONVERSION SYSTEM Applicability Applies to the turbine cycle components for removal of reactor decay heat.
Objective To specify minimum conditions of the turbine cycle equipment necessary to assure the capability to remove decay heat from the reactor core.
Speci fica tions 3.4.1 Thc reactor shall not be heated, above 280F unless the follow-ing conditions are met:
1.
Capability to remove a decay load of 5% full reactor power by at least one of the following means:
a.
A condensate pump and a main feedwater (MFW) pump, using turbine by-pass valve.
b.
A condensate pump and the auxiliary feedwater (AFW) pump using turbine by-pass valve.
2.
Fourteen of the steam system safety valves are operable.
3.
A minimum of 16.3 ft. (107,000 gallons) of water is avail-able in the condensate storage tank.
4 4.
Both emergency feedwater (EFW) pumps and both EFW block valves are capable of automatic actuation, or a dedicated operator is available for their operation.*
5.
Both main steam block valves and both main feedwater iso-lation valves are operable.
6.
The emergency feedwater valves associated with Specifi-cation 3.4.4 shall be operable.
3.4.2 The Steam Line Break Instrumentation and Control System (SLBIC) shall be operable when main steam pressure exceeds 700 psig and shall be set to actuate at 600 + 25 psig.
- One train of EFW may be removed from the control-grade automatic actuation mode for purposes of surveillance testing of the automatic actuation cir-cuitry for a period not to exceed one (1) hour per test without invoking the reporting requirements of Specification 6.12.3.
N#2 40
3.4.3 Components required by Specification 3.4.1 and 3.4.2 to be operable l4 shall not be removed from service for more than 24 consecutive hours.
If the system is not restored to meet the requirements of Specifi-4 cation 3.4.1 and 3.4.2 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the react'r shall be placed I
in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
the requirements of Specification 3.4.1 and 3.4.2 are not met witri.. an additional 4
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 3.4.4 The reactor shall not be heated above 280F unless both EFW pumps are operable.
3.4.5 If the condition specified in 3.4.4 cannot be met:
1.
With one EFW flow path inoperable, the unit shall be brought to hot shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, and if not restored to an operable status within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, the unit shall be brought to cold shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or at the maximum safe rate.
2.
If both EFW trains are inoperable, the AFW pump shall be demonstrated operable immediately, and the unit shall be brought to hot shutdown within one hour.
The unit shall be placed in cold shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or at the maximum safe rate.
3.
If both EFW trains and the AFW pump are inoperable, the unit shall be immediately run back to 15% full power with feedwater supplied from the MFW pumps. As soon as an EFW train or the AFW train is operable, the unit shall be placed in cold shut-down within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or at the maximum safe rate.
l Amendment 4 40a
Bases The feedwater flow required to remove decay heat corresponding to 5% full power with saturated steam at 1065 psia (lowest setting of steam safety valve) as a function of feedwater temperature is:
Feedwater Temperature Flow 60 758 90 777 120 799 140 814 The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 28CF.
Feedwater makeup is supplied by operation of a condensate pump and either a main or the auxiliary feed-water pump.
In the incredible event of loss of all AC power, feedwater is supplied by the turbine driven emergency feedwater pump which takes suction from the condensate storage tank.
Decay heat is removed from a steam generator by steam relief through the atmospheric dump valves or safety valves.
Four-
'een of the steam system safety valves will relieve the necessary amount of ste&m for rated reactor power.
The minimum amount of water in the condensate storage tank would be adequate for about 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of operation.
This is based on the estimate of the aver-age emergency flow to a steam generator being 390 gpm.
This operation time with the volume of water specified would not be reached, since the decay heat remeval system would be brought into operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less.
If the turbine driven emergency feedwater pump has not been verified to be operable within 3 months prior to heatup its operability will be verified upon reaching hot shutdown conditions.
4 The SLBIC System is designed to isolate the steam generators to assure that only one steam generator will experience uncontrolled blowdown following a steam line break.
fiormal steam line operating pressures are approximately 900 psig at all power levels, thus operability above 700 psig with actuation at 600 +25 psig are appropriate.
The setpoint is based on severe transients in the Hiain 5 team lines resulting in rapid pressure decays.
The control-grade EFW automatic actuation system is required per fiUREG-0578 to assure that EFW will be available when necessary.
This control-grade sys-tem is fully testable, but only at the risk of cold EFW reaching a hot steam generator during operation. To reduce the risk of this, and the subsequent transient, the EFW train to be tested may be removed from the automatic actua-tion mode if the other train is operable by automatic action, and the train to be tested is still operable is the manual mode.
References FSAR, Section 10 4l Amendment 4
Table 4.1-1 Instrument Surveillance Requirements Channel Description Check Test Calibra te Remarks 1.
Protective Channel NA M
NA Coincidence Logic 2.
Control Rod Drive NA M(1)
NA (1) To include tripping of breakers via Trip' Breaker shunt trip circuit.
3.
Power Range Amplifier NA NA T/W(1)
(1) Heat balance calibration twice weekly under steady state operating condi-tions, daily under non-steady state operating conditions.
4.
Power Range Channel S
M M(1)(2)
(1) Using incore instrumentation.
M(1)
(2) Axial offset upper and lower chambers monthly and af ter each startup if not donc previous week.
O S.
Intermediate Range Channel S
P/M NA 6.
Source Range Channel S(l)
P NA (1) When in service.
7.
Reactor Coolant Tempera-S M
R ture Channel 8.
High Reactor Coolant S
M R
Prer.sure Channel 9.
Low Reactor Coolant S
M R
Pressure Channel 10.
Flux-Reactor Coolant Flow S
M R
Comparator 11.
Reactor Coolant Pressure S
M R
12.
Pump Flux Comparator S
M R
Table 4.1-1 (Cont'd)
Channel Description Check Test Calibrate Remarks 5
30.
Decay Heat Removal S(l)(2)
M(1)(3)
R (1) Includes RCS Pressure Analog Channel System Isolation Valve
=
Automa~ tic Closure and g
Interlock System (2) Includes CFT Isolation Valve Position (3) Shall Also Be Tested During Refuel-ing Shutdown Prior to Repressuriza-tic t at a pressure greater than 300 but less than 420 psig.
- 31. Turbine Overspeed Trip NA R
f1A Mechanism
- 32. Steam Line Break W
Q R
D Instrumentation and Control System Logic Test & Control Circuits
- 33. Diesel Generator M
Q NA Protective Relaying Starting Interlocks And Circuitry
- 34. Off-site Power Undervoltage W
R R
And Protective Relaying Interlocks And Circuitry
- 35. Borated Water Storage W
NA R
Tank Level Indicator
- 36. Reactor Trip Upon Loss flA PC NA of Main Feedwater Cir-cuitry
Table _4.1-1 (Con _t' d).
fr e
Channel Description Check Test Calibra te Remarks S
- 37. Boric Acid Addition Tank 2-P
- a. Level Channel flA f1A R
fo
- b. Temperature Channel M
flA R
S
- 38. Deleted
- 39. Sodium Hydroxide Tank f4A f1A R
Level Indicator
- 40. Incore Neutron Detectors M(1) flA ftA (1) Check Functioning
- 41. Emergency Plant Radiation M(1) flA R
(1) Battery Check Instruments
- 42. Reactor Trip Upon f1A PC fiA Turbine Trip Circuitry
- 43. Strong Motion Acceleographs Q(1) f1A Q
(1) Battery Check
- 44. ESAS Manual Trip Functions
- a. Switches & Logic fiA P
flA
- b. Logic fin M
f1A
- 45. Reactor Manual Trip f1A P
NA
- 46. Reactor Building Sump Level NA NA R
Table 4.1-1 (Cont'd)
E Channel Description Check Test Cal ibra te Remarks _
k 47.
EFW Actuation Control Logic NA M
R 48.
EFW Flow Indication R
NA R
Note:
S-Each Shift T/W-Twice per Week R-Once every 18 months D-Daily B/M-Every 2 Months NA-Not applicable W-Weekly Q-Quarterly PC-Prior to Going Critical if Not Done Within Previous 31 Days M-Monthly P-Prior to Each Startup if Not Done Previous Week
.