ML17212B291

Emergency Procedure 2-0700040,Revision 0, Loss of Feedwater or Steam Generator Level

ML17212B291
Person / Time
Site:	Saint Lucie
Issue date:	10/12/1981
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17212B290	List: ML17212B289 ML17212B291 ML17212B292 ML17212B293 ML17273A036
References
2-0700040, 2-700040, NUDOCS 8201130426
Download: ML17212B291 (22)

Text

CQ Emergency Procedure 2-0700040 Rev 0

LOF FLORIDA POWER 6 LIGHT COMP ST.

LUCIE PLANT UNIT 2 EMERGENCY PROCEDURE NESTLER 0700040 REVISION 0

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LUCIE PLANT UNIT 2 EMERGENCY EROCED URE NUMBER 2-0700040 REVISION 0

l. 0 SCOPZ This procedure is provided to assist the control roan operators in evaluating and controlling a loss of feedwater or S/G level condition.

NOTE:

Any or all of the following symptoms may be present.

2. 0 SYMPT(MS 2.1 Low-low S/G level 2.1 Indications LIC-9013-A,B,C,D,(RTGB202)

LIC-9023-A,B,C,D,(RTGB202)

LR-9011,9021 (RTGB202).

LIA-9012,9022 (RTGB202)

LR-9012,9022 (RTGB202)

LR-9013D,9023D (RTGB206)

RPS-CH7 2.2 Low feed pump suction or discharge pressure

2. 1 Alarms G-I,G-9,L-11

2. 2 Indica tions PI-1 2-1 9 PIS-09-5 Abnormal Readings On:

AM615/620 AM60S /606 AM62S/626

2. 2 Alarms G-18,G-19,G-1 7 2.3 Incorrect feedwater regulator position (open with high level closed with low level) 2 3 Indications Fl-09-1A, 1B FIC-9011, 9021 FR&9-1A, 1B FCV-9011,9021 2.3 Alarms G-26,G-2 7

EMERGENCY PROCEDURE NUMBER 2-0700040 RE VIS ION 0

2.0 SYMPTCNS

(Cont. )

2.4 Low condensate pump discharge pressure

2. 4 Indications Pl-l2-1 9 2.5 S/G Feed pump trip Alarms G-18,G-19 2.5 Indications Green Lite AM-615 AM-620 2.6 Condensate pump trip

2. 5 Alarms G-l,G-2, G-3 2.6 Indications Green Lite AM-605 AM-606 2; 7 Heater drain pump trip 2-6 Alarms 2.7 Indications Green Lite AM-625 AM-626

2. 7 Alarms F-35, F-3 6 2.8 Increasing T ave*

2.8 Indications TR-1 12QE, TR-1 111, TR-1 121

• Initial increase, then decrease 2.8 Alarms K-17

EMERGENCY PROC ED URE N UMBER 2-0 700040 REVIS ION 0

2. 0 SYMPTOMS:

(Cont. )

2-9 Increasing RCS Pressure (initially) 2.9 Indications PI 1102A, B,

C, D

PR-1 100 P IC-110X, Y

RPSWH5

2. 9 Alarms L-20, L-28 NOTE: If a feedwater break occurs inside containment, be especially alert for erroneous indications.

2.1.0 Increasing containment pr essur e 2.10 Indications Break inside containment PIS-07-2A, 2B, 2C, 2D (RTGB206)

PI-07-4A,SA (RTGB206)

PR-07-4B,SB (RTGB20 6) 2.10 Alarms L-5,L-13, Q-3, P-3, P-1 3, P-2 3

2. 11 Increasing containment temperatur e 2 ~ 11 TI-07-3A, 5A TR-0 7-3B, 5B 2.11 Alarms (RTGB206)

(RTGB 206) 2.12 Increasing reactor cavity sump level 2.13 Feed flow/steam flow mi sma tch

2. 12 Indica tions LISW 7-6 (RTGB 205)

FR-07-3 (RTGB205) 2.12 Alarms N-29,N-21,N-4 6 2.13 Indications FR-8011,9011 FR-8021,9021 FR-09-1A, 1B

2. 13 Alarms

LOF EMERGENCY PROC ED URE NUMBER 2-0 700040 REVIS ION 0

2. 0 SYMPTOMS:

(Cont. )

2. 14 Reactor Trip 2.14 Indications CEA's Inserted (ADS)

RPS-CH.7 tripped RPS-RTCB's Open 2.14 Alarms L-3,L-11, K-l,K-2,K-3 K-4,K-5,K-9,K-1 0 A

2.15 Turbine Trip

2. 15 Indications WM881-0 Magawatts Meter for generator output 9 zero W-REC-871

2. 15 Alarms D-8,C-3 2.16 After Trip, following key paraneters:

2.16. 1 Reactor Power-Decrease

2. 16. 2 Przr Pr essur e-Initial

increase, then decrease, then stabilize

2. 16. 3 Tempe ratur e-initial

increase, then decrease 2.16.4 Pressurizer level slight

increase, then decrease 2.16.5 S/G Pressure slight

increase, then decrease 2.16.6 S/G Level-decrease NOTE: Break Diagnostic Chart Fig.

1 may assist in evaluating the situation.

EMERGENCY PROCEDURE NUMBER 2-0700040 RE VIS ION 0

LOF

Cl

EMERGENCY PROCED URE NUMBER 2-0 700040 RE VIS ION 0

3 0 AUTCNATIC ACTIONS 3.1 Reactor Trip (Low S/G level) 3 1.1 3.1.2

3. 1.3 3.1.4 Reactor trip breakers open CEA's insert Reactor Power decreasing AFV Start (t 34K Level 3.2 Turbine Trip (Reactor trip turbine) 3 ~ 2.1 3 ~ 2 ~ 2 3 ~ 2 ~ 3

3. 2-4 Turbine valves close Generator breakers open Auxiliary power transfers to Startup transformers Main Pdwtr Valves close, bypasses open to 5X flow position

3. 3 Turbine Runback (los s of one feedwa ter pump) 3 3.1 30 30 2 Load reduction

> 40 of rated power Operation of steam bypass control system and/or atmospheric dumps

4. 0 IMMEDIATE OPERATOR ACTIONS

4. 1 Ensure imaediate operator actions for a reactor trip have been accanplished.

4.1.1 Any autoaetic functions that were required have operated properly.

4. 2 A feedwa ter b reak is indicated

4. 2 Continuously decreasing S/G level and pressure with increased fdwtr flow-THEN Isolate the break with controls available in the control room Isolate the break manually with local controls.

4-3 Continue feeding the non-faulted S/G with the available source.

4.3 Either main fdwtr or aux fdwtr 4.4 If possible maintain one condensate pump running 4.5 Verify proper operation of przr. pressure and level control systems 4-4 4

5 To provide secondary system system cooling.

Maintain primary system control.

Cl

LOF EMERGENCY PROCED URE NUMBER 2-0700040 REVIS ION 0

4.0 IMMEDIATE OPERATION ACTIONS:

(Cont. )

If a non-isolable break has

occurred, immediately refer to Main Steam Line Break procedure, EP 1-0810040.

4-6 If the following symptoms are indicated, refer to ICC procedure 2-0120043 a

S/G level at or near zero NOTE:

Loss of feedwater to both steam generators during power operation has the potential of producing conditions which could lead to inadequate core cooling'.

S/G pressure rises to dump valve setpoint, unless a break occurs in which case pressure would be continually decreasing NOTES:

Following an SIAS caused by low RCS pressure and after it has been verified that CEA's have been fully inserted for 5 seconds, stop all operating RCP s

c.

Coincident pressure in the S/G and the and the primary system subcooled initidly and decrease, then increase sharply e.

After reaching approx constant temperature, the primary coolant temperature increases well above secondary saturation temp Following any SIAS, operate the SIS until RCS hot and cold leg temperatures are at least 50oF below saturation temperature for RCS pressure and a pressurizer level is indicated, unless the cause of the SIAS has been verified to be an inadvertent actuation.

If 50oF subcooling cannot be maintained after the system has

stopped, the HPSI system must be restarted.

Reactor Coolant Pumps may be restarted if it has been confirmed that a LOCA has not occurred, pump services can be restored, and RCS pressure-temperature conditions permit restart.

1

LOF EMERGENCY PROC ED URE NUMBER 2-0 70004 0 REVIS ION 0

5.0 SUBSE UENT ACTIONS 5.1 Refer to Reactor Trip/Turbine Trip, 2-0030130 and ensure that all subsequent actions (Section 5) have been or are being performed.

CHECK 5.2 If partial loss of feedwater results in a turbine runback:

5. 2.1 Verify the load reduction is

> 40X of rated pwr.

5.2.2 Attempt to restore main feedwater using manual control.

5.2.3 Maintain S/G levels with the remaining feedwater source.

5.2.4 Match turbine power with reactor power (T avg Tref), within the capacity of the remaining feedwa ter source.

5.3 If both feed pumps are lost:

5.3. 1 Initiate aux feed 5.3.2

~Sto (2) Heater Drain Pumps 5.3.3

~Sto (1) Condensate Pump 5.3-4

~Sto (1)

RCP in each loop 5.4 Close S/G blowdown isolation valves 5-4.1 Close FCV-23-3, 5, 4, 6

5.4.2 Close FCV-23-7, '9 CAUTION:

Overfeeding the S/G s can cause excessive cooldown

(> 75oF/hr)

Cl

LOF EMERGENCY PROCED URE N UMBER 2-0 700040 REVIS ION 0

5-0 SUBSE UENT ACTIONS:

(Cont. )

5.5 If the motor driven aux fdwtr pp capability is lost, perform the following:

CHECK 5.5. I Open MV-08-13 (2A) OR MV-08-1 4 (2B)

Stean Supply Valve to aux fdwtr pp.

5.5.2 Control turbine speed with MV-08-3 for the desired rate.

5.5.3 Control feed flow to S/G's with MV-09-11 (2A S/G)

MV-09-12 (2B S/G) 5-6 For manual operation of AFW.

5. 6. 1 Es tab lish canmun ications between the control room and AFW station.

5.6.2 If local manual control is needed, OPEN breakers 2-50308 and 2-60320 8 2AB-125VDC bus.

5.6.3 Throttle MV-08-3 to raise 2C aux fd pp discharge pressure to approx 1200 psig.

5.7 If S/G level is < 38% do not exceed 785 GEf.

(Maximum flow rate of 2 elect AFW Pumps)

(Maximum flow rate. of Steam AFW Pumps) 5.8 Check auxiliary feedwater flow utilizing all available indications (flow indicators, flow recorders, S/G levels)

CAUTION:

Do not feed a dry S/G unless both S/G become dry.

Then admit fdwtr to only one S/G.

5.9 Observe CST level frequently (1/30 min) 5.10 If condenser vacuum is lost or MSIV s close, take manual control of the atmospheric dumps for S/G pressure control.

LVi EMERGENCY PROC ED URE N UMB ER 2>>0 700040 RE VIS ION 0

5.0 SUBSE UENT ACTIONS:

(Cont.)

5.11 Ensure that the pressurizer pressure control system is automatically restoring RCS pressure.

CHECK 5.12 Ensure that the CVCS is restoring pressurizer level, (MAX charging, MIN letdown)

~

NOTE:

Any system that is placed in manual, requires more operator attention 5.13 Coordinate operations with Unit 81 to assure sufficient condensate

via, 5.13.1 Water treatment plant and/or 5.13.2 City Wtr Str Tk and/or 5.13.3 Unit Fl CST 5.14 Folio~ appropriate procedures for cooldown or return to power.

Do not exceed 75oF/hr cooldown rate.

EMERGENCY PROC ED URE N UMBER 2-0 70004 0 REVIS ION 0

6 ~ 0 PURPOSE/DISCUSSION 6.1 This procedure provides instructions to be followed in the event of loss of feedwater flow and/or loss of steam generator level. It is to be used when specific symptoms indicate this condition and the immediate actions of the Reactor Trip/Turbine Trip procedures have been ccmpleted.

N 6.2 Loss of stean generator level results in a reduction in capability of the secondary system to remove the heat generated in the reactor core.

Loss of stean generator level results from the inability to supply feedwater in an amount equal to the existing steam demand.

Excessive steam demand, i.e., above feedwater system capability is considered in Emergency O.P.

2-0810040, Main SE ncn Line Break.

6.3 The loss of all feedwater to the steam generator and the subsequent heatup of the Reactor Coolant System will result in saturation conditions at high temperature and pressure in the Reactor Coolant System.

This would cause a loss of reactor coolant mass through the PORV's and code safeties and void formations throughout the system.

A maximum effort should be directed toward the initiation of feedwater flow to the steam generators-Without feedwater, the steam generators could boil dry within approximately 13 minutes.

6.4 The inability to supply feedwater in the required quantity could result from one or a combination of the following:

6.4-1 Pipe break in the condensate/feedwater system

~

6.4.2 Flowpath blockage due to valve closure or strainer stoppage.

6.4.3 Loss of pumping capacity due to pump trip, loss of AC power, or pump cavitation.

6.4-4 Loss of 2C Auxiliary Feedwater pump remote operating capability-6.5 The following protective functions are provided to prevent loss of heat sink.

a)

Reactor trip:

Stean generator low-low level at 39/.

EMERGENCY PROCEDURE NUMBER 2-0700040 REVIS ION 0

6.0 PURPOSE/DISCUSSION:

(Cont.)

6. 5 (Cont. )

b)

Turbine Runback; due to:

1.

Loss of both heater drain pumps if turbine power is

> 92X.

IIORlf 2.

Loss of feedwater pump if turbine power is

> 60X.

c)

AFH Pump Auto-start at 34X steam generator level.

6.6 The action taken for loss of steam generator level is basically the same regardless of cause.

That is, to trip the reactor, thereby removing the heat source, and restoring steam generator level with the auxiliary feedwater pumps.

This approach is due to the self-sufficient nature of the auxiliary feed system.

It, in no way, depends on the normal feedwater systems The auxiliary feed system has the capability to bring the unit to hot shutdown and remain there for a period in excess of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, regardless of any single failure in the normal feed system.

Three auxiliary feed pumps are available.

Wo motor driven pumps, either of which can supply either steam generator and one steam driven pump which can supply either stean generator.

The auxiliary feedwater system ties into each steam generator feed line downstream of the steam generator feed inlet check valve.

Capability exists to feed either steam generator by any auxiliary feed pump.

A break in the feed line between the feed line check valve and the steaa generator would result in the loss of that steam generator as a heat sink-Corrective action would be to isolate feed flow to that steam generator and continue maintaining level in the unaffected steam generator.

Residual heat can be adequately dissipated in this manner.

The leak must be isolated in any incident involving a break even though the break does not immediately affect the auxiliary feedwater system.

The isolation of a leak is necessary to conserve the water available in the condensate storage tank.

The leak is isolated by stopping the steam generator feed pumps and condensate pumps.

A leak downstream of a steam generator feed line check valve would also require stopping auxiliary feed flow to that steam generator.

LOF EMERGENCY PROCED URE NUMB ER 2-0 700040 REVIS ION 0

6. 0 PURPOSE/DISCUSSION:

(Cont ~ )

6.7 The Aux feed pumps will auto start approximately 3 minutes following either S/G level dropping below 34X (2/4 logic) ~

Flow is automatically initiated to both steam generators and cannot be terminated for 30 seconds.

Bypass switches are provided on RTGB 102 which will prevent pump starts and valve stroke and would be utilized if normal feedwater flow remains or becomes available and is desired.

7. 0 REFERENCES 7.1 FSAR, Section 15.2.8 and Section 10.5.3 7.2 Combustion Engineering Emergency Procedure F-EP-9

7. 3 C.E-LOF guidelines, CEN>>l 28 7.4

'78%L EP 0700040 Rev.

8 (Unit 1)

('as of Feedwater or Steam Generator Level) 7.5 Draft of NUREG 0799 8 0 RECORDS R

UIRED Log Book Entries Applicable chart recorders

9. 0 AP PRO VAL Reviewed by the Facility Review Group Approved by Plant Manager Rev ~

Reviewed by FacQ.ity Review Group Approved by Plant Manager 19 19 19 19 "LAST'PAGE" Emergency Procedure 2-0700040 Rev. 0 TOTAL NO.

PAGES 14