ML19318D237

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Emergency Operating Procedure, Steam Generator Tube Rupture, Vol 5,Revision 0
ML19318D237
Person / Time
Site: Farley Southern Nuclear icon.png
Issue date: 05/14/1980
From:
ALABAMA POWER CO.
To:
Shared Package
ML19318D229 List:
References
FNP-2-EOP-3.0, NUDOCS 8007080236
Download: ML19318D237 (9)
v • d • e


Text

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O VOLUME 5 FNP-2-EOP-3.0 ..

May 14, 1980 Revision 0 FARLEY NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE ,

FNP-2-EOP-3.0 S

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REVISION NO. 0 1

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VOLUME 5 FNP-2-EOP-3.0 1

FARLEY NUCLEAR PLANT -

UNIT 2 EMERGENCY OPERATING PROCEDURE FNP-2-EOP-3.0 .

STEAM GENERATOR TUBE RUP*xURE 1.0 Purpose _

The purpose of this procedure is to describe the required actions following a steam generator tube rupture accident. The objectives of these actions are as follows:

1.1 To minimize the release of radioactive material to the environment.

1.2 To maintain the ability to remove residual heat from the reactor coolant system.

1.3 To prevent carryover of water from the affected steam generator into the main steam system.

2.0 Symptoms

.I Refer to Section 2.3 of FNP-2-EOP-0.

3.0 Automatic Actions Refer to Section 3.3 of FNP-2-EOP-0. l 1

4.0 Immediate Operator Actions

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Refer to Section 4.0 of FNP-2-EOP-0.

5.0 Subsequent Operator Actions CAUTION No single plant parameter should be relied upon for evaluating the plant status under accident conditions.

NOTE The process variables referred to in this procedure are typically monitored by more than one instrumentation channel. The redundant channels should be checked for consistency while performing the steps of this procedure.

5.1 Actuate the PLANT EMERGENCY ALARM and announce over the public address system the following:

ALL PERSONNEL REPORT TO YOUR DESIGNATED ASSEMBLY AREAS. .

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s VOLUME 5 FNP-2-EOP-3.0 5.2 Monitor the fcilowing: -

RCS pressure pressurizer level wide range RCS temperature BIT flow steam generator pressure steam generator water level auxiliary feedwater flow to steam generators CTMT pressure accumulator levels RER flow RWST level CTMT spray flow spray additive tank flow Margin to Saturation (PSIA) to assess the nature and severity of the accident and to verify proper operation of safety systems.

CAUTION System pressure affects the accuracy of level indicating instruments. Steam generator level should not be used as sole indication of adequate

. leve1 in the S/G to provide a heat sink nor should I pressurizer level be used as sole indication of proper pressurizer level. Parameters such as AUXILIARY FfwnWATER FLOW, STEAM PRESSURE, RCS PRESSURE, response of RCS PRESSURE to pressurizer heater operation, and RCS WIDE RANGE T and T should also be evaluated. 'The minimum hand c maximum allowed levels necessary to ensure that actual level is between the instrument's lower tap and upper tap are shown for various pressures in table 1.

5.3 If any ESF actuating parameters exceed their 1

actuation setpoint without actuation, manually initiate that ESF system.

1 5.4 Identify the affected steam generator by the following:

5.4.1 Steam generatar level and pressure indicators and recorders.

5.4.2 Feed flow / steam flow recorders.

5.4.3 If uncertainty exists, all steam generators should be sampled for activity.

5.5 Initiate EIP-9 Radiation Exposkre Estimation and Classification of Emergencies.

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VOLUME 5 FNP-2-EOP-3.0 5.6 When the affected steam generator has been posi- -

tively identified, ISOLATE THE AFFECTED STEAM GENERATOR by closing or verifying closed the following:

5.6.1 Atmospheric relief valve for the affected steam generator, place in the manual mode.

5.6.2 MSIV's for the affected steam generator.

5.6.3 MSIV bypass valve 2-MS-HV-3976A (B,C).

5.6.4 Main feedwater stop valve 2-CFW-MOV-3232A (B,C).

5.6.5 Auxiliary Feedwater stop valve 2-AFW-MOV-3764A and E (B and D, C and F) ,

5.6.6 TDAEVP steam supply from S/G 2B(C) valve 2-AFW-HV-3235A (B).

5.7 If RCS pressure decreases to less than 1300 psig (other than controlled RCS depressurization),

trip all Reactor Coolant Pumps after the high F.

head safety injection pump operation has,been ^

verified.

NOTE If reactor coolant pumps are tripped, frequently monitor WR RCS hot leg temperature, WR RCS cold leg temperature, and incore thermocouples to verify circulation and core cooling as indicated

_ by:

The AT between WR RCS hot leg and cold leg <64 0F in the unaffected loops, or Incore thermocouples at or near RCS WR hot leg temperature in the unaffected loops and decreasing.

If these conditions do not exist, maintain Auxiliary Feedwater flow rate and increase steam dump rate to enhance the establishment of natural circulation.

NOTE: The conditions listed above may be continuously monitored by use of the AT function on the Core S2bcooling Monitoring Panel.

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C VOLUME 5

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.. FNP-2-EOP-3.0 j 5.8 RESET the SAFETY INJECTION and CTMT PHASE "A"

  • ISOLATION signals.

l CAUTION Automatic reinitiation of safety injection will not occur since the reactor trip breakers are .

not reset.

5.9 Control Auxiliary Feedwater flow to the unaffected steam generators as necessary to maintain no-load S/G level of 33%.

5.10 Open turbine building train A and B Service Water isolation valves Q2P16V514; 515; 516; 517.

5.11 Stop all charging pumps not required to maintain pressurizer level greater than 20%. i 5.12 Stop RER pumps A and B.

l 5.13 After the affected steam generator has been identified and isolated, begin a rapid cooldown of the reactor coolant system to approximately 497'F as follows:

f 5.13.1 With the main condenser availablF, transfer steam dump control to the Steam Pressure Mode and dump steam to the main condenser.

5.13.2 With off-site power or main condenser not available, dump steam through the non-affected steam generator atmospheric rela.ef valves.

5.14 After the reactor coolant system temperature has been reduced to approximately 497*F, begin a depressurization of the reactor coolant system to a value equal to the affected steam generator steam pressure as follows:

5.14.1 -If the RCP's are in service, use the pressurizer spray to reduce the pressure.

5.14.2 If off-site power is not available, or the RCP's are not in service, open one 4

pressurizer power operated relief valve to reduce reactor coolant pressure.

Monitor the pressurizer relief tank pressure and do not exceed 100 psig in the PRT. A 4 Rev. O

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VOIUME 5 FNP-2-EOP-300 oj NOTE .

If the pressurizer power operated relief valve is opened, an increase in pressurizer level may occur as liquid replaces the escaping steam. Do not allow the pressurizer level to increase above 50%.

5.15 Re-establish pressurizer heater control to maintain reactor coolant pressure. If an LOSP has occurred, re-establish pressurizer heater control per FNP-2-EOP-7.0, Section 5.3.

5.16 Establish normal charging and letdown control to maintain 20% pressurizer level as follows:

5.16.1 Stop all but one charging pump.

5.16.2 Open letdown line CTMT isolation valve 2-CVC-HV-8152.

5.16.3 Open letdown line iso valves 2-CVC-HV-459 and 460.

5.16.4 Open letdown orifice valve 2-CVC-HV-8149B(C).

I--' 5.16.5 Open charging pumps to regenerative heat exchanger valves 2-CVC-MOV-8107 and 2-CVC-MOV-8108.

5.16.6 Open charging pump A, B and C miniflow valves 2-CVC-MOV-8109 A, B and C.

5.16.7 Open charging pumps miniflow isolation valve 2-CVC-MOV-8106.

5.16.8 Close boron injection tank inlet valves 2-CVC-MOV-8803A and B.

5.16.9 Place charging flow control valve 2-CVC-FCV-122 in AUTO.

5.16.10 Verify VCT level greater than 14% and reactor makeup in automatic with blended flow set for the hot shutdown boron concentration.

5.16.11 Open VCT outlet isolation valves 2-CVC-LCV-115C and E.

5.16.12 Close RWST to charging pump header valves 2-CVC-LCV-115B And D.

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VOLUME 5 FNP-2-EOP-3.0

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5.16.13 Open RCP seal water return isolation ..

valves 2-CVC-MOV-8100 and 2-CVC-MOV-8112.

5.17 If off-site power is reliable as verified by the availability of two independent off-site sources, then shutdown the diesel generators and align for automatic operation per SOP-38.0.

5.18 Continue cooldown in accordance with UOP-2.2 by use of steam dumps or by dumping steam through the non-affected steam generators' atmospheric relief valves if main condenser is not available.

Cooldown.as rapidly as possible within cooldown limitations, and maintaining NPSH requirement 4

for Reactor Coolant Pumps. Maintain RCS pressure less than the affected steam generator steam pressure.

NOTE The boron concentration in the reactor coolant must be verified before additional cooldown is initiated and checked periodically to detect any dilution effect caused by leakage of secondary water into,the reactor coolant.

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Throughout the cooldown, maintain a steam bubble in the pressurizer. Solid water pressure control may not be effective.

5.18 Monitor the condensate Storage Tank level. If the Condensate Storage Tank level decreases to fcur (4) feet, and auxiliary feedwater is required by plant conditions, shift the auxiliary feedwater pumps suction to the Service Water System per FNP-2-SOP-22.0, Section 4.7.

5.19 Return systems that are not required to be aligned for the accident condition to normal operation, refer to AOP-4.0.

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  • FNP-2-EOP-3.0 Rsv. O TABIZ 1 --

SUBC00IID CONDITIONS SATURATED STEAM _ ,

RCS PRESSURE MAX. TH-W.R. (*F) TH-W.R. RCS PRESSURE

. .. (PSIG) 50' (*F) '

(PSIG) 0 162 _

50 248 220 3

_ 100 288 230 6 150 316 240 10 200 338 25 0 , 15 25 0 356 260 21 300 372 270 27 350 386 280 35 400 398 290 43 450 410 300 52 500 420 310 63 550 430 320 75 600 439 330 88 650 447 340 103 700 456 350 120 750 463 360 138 800 470 370 159 850 .

477 380 181 900 484 390 206 950 490- 400 233

'[_ 1000 496 410 ~262 1050 502 420 294 1100 508 430 329 1150 514 440 367 1200. 519 450 408 1250 524 460 452 1300 529 470 500 1350 534 480 551 1400 539 490 607 1450 543 500 666 1500 548 510 730 1550 552 520 798 1600 556 530 871 1650 560 540 948 1700 564 550 1031 1750 568 560 1119 1800 572 570 1212 1850 576 580 1311 1900 580 590 1417 1950 583 600 1529 2000 587 610 1647 2050 590 620 1772 2100 594 630 1905 2150 597 640 2046 2200 600 650 2194 2250 604 660 2352 4

2300 607 670 2518 2350 610 680 2698 2400 613 690 2880

'2450 616 700 3079

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