ML19210E367

Operating Guidelines for Small Breaks

ML19210E367
Person / Time
Site:	Rancho Seco
Issue date:	11/30/1979
From:	Beckner D, Hallman D, Kaurank B BABCOCK & WILCOX CO.
To:
Shared Package
ML19210E356	List: ML19210E355 ML19210E367 ML19210E369
References
BWNP-20004, NUDOCS 7912040412
Download: ML19210E367 (49)

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BWNP-20004 (6-76)

BABCOCK & WILCOX NUCLEAR POWEE CENERAflCN DIVI $lCN 1ECHNICAL DOCUMENT PRMIHRY EMERGENCY OPERATING SPECIFICATION 69 1106001 - 00 Doc.10 - Serial No., Revision No.

for OPERATING GUIDELINES FOR SMALL BREAKS FOR OCONEE 1, 2 AND 3, THREE MILE ISLAND 1, 2, CRYSTAL RIVER 3, AND RANCHO SECO 1 1473 015 PAGE 1

BWNP-20005 (6-76)

BABCOCK & WILCOX NUM 8tR NUCLEAR POWit GENERATION DivillON RECORD OF REVISION 69-1106001-00 REY. N0.

CHANGE SECT / PARA.

DESCRIPTION / CHANGE AUTHORIZATION 00 Original Issue - D. A. Beckner Customer Services Iw*w DATE //!

7 PREPARED BY L Mp

/7 APPROVED BY A_

qr3 DATE[r

//!Z[79 APPROVED BY Mey b/w DATE (Title)J 7

(Name)"

//ukd um 8CSDATE

1. /2,[77 APPROVED BY m

APPROVED BY DATE

/ 2 7f (Name)

(T ple) e 1 A73 0%

DATE' 11-1-7 PAGE 2

69-1106001 PART 1 - OPERATING CUIDELINES FOR SMALL BREAKS 10 SYMPTOMS A':D INDICATIONS (US!EDIATE INDICATIONS) 1.1 EXCESSIVE REACTOR COOLANT SYSTD1 (RCS) MAREUP*

1.2 DECREASING RCS PRESSURE 1.3 REACTOR TRIP 1.4 DECREASING PRESSURIZER LEVEL

• 1.5 ESFAS ACTUATION

• 1.6 LOW MAKEUP TANK LEVEL

• 1.7 ADDITIONAL CRITERIA DURING HEATUP AND C00LDOW

• 1.7.1 RCS TDfP INCREASING, MINIMUM LETDOWN AND PRESSURIZER LEVEL DECREASING 1.7.2 WITH A C00LDOWN OF $ 100*F/HR AND CANNOT MAINTAIN LEVEL IN MAKEUP TANK

• MAY NOT OCCUR ON ALL SMALL BREAKS 2.0 DDEDIATE ACTIONS o

2.1 IF THE ESFAS HAS BEEN INITIATED AUTOMATICALLY BECAUSE CF LOW RC PRESSURE, DSEDIATELY SECURE ALL RC PUMPS.

2.2 VERIFY CONTROL ROOM INDICATIONS SUPPORT THE ALARMS RECEIVED, VERIFY AUTOMATIC ACTIONS, AND CARRY OUT STANDARD POST-TRIP ACTIONS.

2.3 3ALANCE HICH-PRESSURE INJECTION (HPI) FLOW BETWEEN ALL INJECTION LINES WHEN HPI IS INITIATED.

o 2.4 VERIFY THAT APPROPRIATE ONCE-THROUGH STEAM GENERATOR (OTSG) LEVEL IS MAINTAINED BY FEEDWATER CONTROL (LOW LEVEL LDiIT WITH RC PGES OPERATING, DERGENCY LEVEL WITHOUT RC PL?TS OPERATING).

o 2.5 MONITOR SYSTEM PRESSURE AND TEMPERATURE.

IF SATURATED CONDITIONS OCCUR, INITIATE HPI.

2.6 IF ESFAS HAS BEEN BYPASSED DUE TO HEATUP OR C00LDOWN, INITIATE SAFETY INJECTION.

CAUTION:

IF 50*F SUSC00 LING CRITERIA IS MET, THROTTLE HPI FLOW TO KEEP SYSTD1 PRESSURE WITHIN NORMAL TECHNICAL SPECI-FICATION P-T CURVE LD!ITS.

IF RCS IS NOT 50*F SUBC00 LED, CONTINUE FULL SAFETY INJECTION UNTIL 50*F SUEC00 LING IS ATTAINED OR THE P-T LDiITS OF FIGURE 1 ARE REACHED.

)k7b b 7 69-1106001 3.0 PRECAU IONS o

3.1 IF THE ESFAS HAS BEEN INITIATED ON LOW RC PRESSURE, TE?3fI::ATIC : OF RC PL"!P OPERATIO : TAKES PRECEDENCE OVER ALL OTHER UCIEDIATE ACTIONS.

NOTE:

IF ESFAS HAS BEEN ACTUATED ON HIGH R3 PRESSURE, THEN MONITOR RC PPISSURE AND TRIP RL PD!PS ONCE.RESSURE DECREASES BELOW THE ESFAS LOW PRESSURE SETPOINT.

3.2 IF ESFAS HAS BE!:N INITIATED, TilC RC PUMP'S TRIPPED, AND THE RCS DETERMINED TO BE AT LEAST 50 F SUBC00 LED. THE OPERATOR SHOULD ESTABLISil AS QUICKLY AS POSSIBLE IF Tile CAUSE FOR Ti!E DEPRESSURI:'A-TION IS DUE TO EITi!ER A LOCA OR NON-LOCA (OVERC00 LING) EVENT.

PROCEED TO STEP 4.4 FOR NON-LOCA EVENTS.

3.3 IF THE HPI SYSTD! !!AS ACTUATED BECAUSE OF LOW PRESSUP2 CONDITIONS, IT MUST RDIAIN IN OPERATION UNTIL ONE OF THE FOLLOWING CRITERIA IJ SATISFIED:

1.

THE LPI SYSTDI IS IN OPERATION AND FLOWING AT A RATE IN EXCESS OF 1000 GFM IN EACH LINE AND THE SITUATION HAS BEEN STABLE FOR 20 MINUTES.

OR 2.

ALL HOT AND CcLD LEG TEMPERATURES ARE AT LEAST 50F BELOW THE SATURATION TEMPERATURE FOR THE EXISTING RCS PRESSURE

- AND -

THE ACTION IS NECESSARY TO PREVENT THE INDICATED PRESSURIZER LEVEL FROM GOING OFF-SCALE HIGH.

NOTE:

IF 50F SUBC00 LING CANNOT BE MAINTAINED, THE HPI SHALL 1

f BE REACTIVATED.

NOTE: THE DEGREE OF SUBC00 LING BEYOND 50F AND THE LENGTH OF TIME HPI IS IN OPERATION SHALL BE LIMITED BY THE PRESSURE /

I TEMPERATURE CONSIDERATIONS FOR THE VESSEL INTEGRITY (SEE SECTION 3.4).

g 303 08

_2-

69-1106001 3.4 WHEN THE REACTOR COOLANT IS > 50 F SUBC00 LED, THE REACTOR VESSEL DOWNCOMER PRESSURE /TD!PERATURE (P-T) CO3GINATION SHALL BE KEPT BELOW AND TO THE RIGHT OF THE LU!IT CURVE SHOWN IN FIGURE 1.

THE DOCCOLER TDTERATURE SHALL BE DETER 11NED AS FOLLOWS:

3.4.1 WITH ONE OR MORE RC PUITS OPERATING USE ANY COLD LEG RTD AS AN DIDICATION OF REACTOR VESSEL DOWNCO'ER TDTERATURE.

').4.2 WITH NO RC PU:IPS OPERATING THE RV DOWNCOMER TDTERATURE SHALL BE DETERMINED BY AVERAGIN2 THE FIVE LOWEST INCORE THER10 COUPLE TDTERATURE REA"'..iGS AND SUBTRACTING 150 F FROM THE AVERAGE INCORE THERMOCOUPLE TDTERATURE VALUE.

5 te - 150 F T

=

DW WHERE T

= AVERAGE RV MCO.ER TDMM, F DWN 5

IT

= Sm! 0F THE 5 LOWEST INCORE THERIOCOUPLE TDTERATURE 7c READINGS.

NOTI:;: FIGURE 1 IS APPLICABLE ONLY UNDER LOCA CONDITIONS. 'THE P/T CURVE IN THE TECHNICAL SPECIFICATION IS VALID FOR ALL OTHER OPERATDIG CONDITIONS.

NOTE: WHEN THE REACTOR COOLANT IS LESS'THAN 50 F SUBC00 LED, THE REACTOR VESSEL DOWNCOMER PRESSURE TDTERATURE COMBINATION WILL INHERDITLY BE BELOW AND TO THE RIGHT OF THE LIMIT CURVE.

TIIEREFORE, NO OPERATOR ACTION WILL BE REQUIRED TO PREVENT EXCEEDDIG THE REACTOR VESSEL INTEGRITY LDiITS UNTIL AFTER A > 50 F SUBC00 LED MARGIN EXISTS.

!473 019 Revised

69-1106001 NOTE:

Wi!EN THE REACTOR COOLANT IS R 50*F SUBC00 LED, RC PRESSURE CAN BE REDUCED BY REDUCING THE HPI FLOW RATE TO AVOID ENCEEDING THE RV INTECRITY LU!ITS.

35 PRESSURIZER LEVEL MAY BE INCREASING DUE TO RCS Er. ACHING SATUPJT".D C '_

9NS OR A BREAK ON TOP OF THE PRESSURIZER.

3.6 IF HIGH ACTIVITY IS DETECTED IN A STEA1! GENERATOR, ISOLATE THE LEAKI'G GENERATOR.

IT IS RECom! ENDED THAT BOTH STEAM GENL2ATORS NOT BE ISOLATED.

3.7 OTHER INDICATIONS WHICH CAN CONFIPJi THE D:ISTENCE OF A LOCA:

37.1 RC DRAIN TANK (QUENCH TANK) PRESSURE (RUPTURE DIS!" MAY BE ELOWN).

3. 7. 2 INCREASING REACTOR BUILDING SUMP LEVEL.

3. 7. 3 INCREASING REACTOR P.UILDINC h.1TERATURE.

3. 7. 4 INCREASING PIACTC" BUILDING PRESSURE.

3. 7. 5 INCREASING RADIATION MONITOR READINGS INSIDE CONTAINMENT

3. 7. 6 REACTOR COOLANT SYSTDI TDIPERATURE BECOMING SATURATED RELATIVE TO THE RCS PRESSURE.

3. 7. 7 HOT LEG TDGERATURE EQUALS OR EXCEEDS PRESSURIZER TDIPERATURE.

3.7.8 INCREASE IN THE EXCORE NEUTRON DETECTOR D DICATIONS.

h NOTE:

IN CONJUNCTION WITH TRE INDICATIONS IN 3.10.1, THIS f

COULD BE AN' INDICATION OF INADEQUATE CORE COOLING.

d 3.8 HPI COOLING REQUIRDIENTS COULD DEPLETE THE EORATED WATER STORAGE TANK, AND INITIATION OF LPI FLOW FRCM UIE REACTOR BUILDING SUMP TO.THE HPI PUMPS WeaD BE REQUIRED.

3.9 ALTERNATE INSTRUMENT CHANNELS SHOULD BE CHECKED AS AVAII ABLE TO CONFIDI KEY PARAMETER READINGS (IE, SYSTDi TDTERATURES, PRESSURES AND PRESSURIZER LEVEL).

3.10 MAINTAIN A TDTERATURE VERSUS TDIE PLOT AND A CCRRESPOND' NG TDTERATURE I

PRESSURE PLOT ON A SATURATION DIAGRAM. USD:G EDT LEG RTD'S AND HIGHEST INCORE THERMOCOUPLE READING, THESE PLOTS WILL MAKE IT POSSIBLE TO TRACK THE PLANT'S CONDITION THROUGH PLANT C00LDOWN.

3.10.1 IF EITHER OF THE FOLLOWING INDICATIONS OF INADEQUATE CORE COOLING EXIST, GO TO SECTION 4.5.

1.

HOT LEG RTD'S READ SUPERHEATED FOR THE EXISTING RCS PRESSURE.

2.

INCORE THERMOCCUPLE TD9 READS SUPERHEATED FOR THE EXISTING RCS PRESSURE.

M 3.10.2 IF PRIMARY TDiPERATURE AND PRESSURE IS Ut. CREASING ALONG THE SATURATION CURVE THEN SU3C00 LED CONDITIONS WILL 3E ESTABLISNED.

THIS WILL BE INDICATED BY PRDIARY SYSTD1 PRESSURE NO LONGER FOLLOWING THE SATURATION CURVE, AS PRIMARY SYSTD1 TDfP. DECREASES.

IGIEN THIS OCCURS, PRD!ARY SYSTDi PRESSURE SHOULD BE CONTROLLED

!473 020

_4

69-1106001 THE BY ADJUSTING HPI FLOW, TO man;TAIN 50*F SUBC00 LING.

DEGREE OF SUBC00 LING BEYOND 50*F SHALL BE CONTROLLED WITHIN THE LIMITS DEFINED IN SECTION 3.4.

3.11 COMPONENT COOLING WATER (CCW) AND S5'AL INJECTION SHOULD B THE RC PUMPS TO INSURE CONTDIUED SERVICE OR THE A3ILITY TO P.ESTAR PUMPS AT A LATER TD'E.

3.11.1 IF CCW IS LOST A:m THE RC PUMPS ARE OPERATIVE. CCU MUST BE o.

RESTORI:D WIT!!IN 30 MINUTES OR THE RC PUMPS MUST BC MANUALLY TRIPPED.

3.11.2 IF THE RC PUMPS ARE TRIPPED FOR ANY REASON, SEAL INJECTION o

SHOULD BE MAINTAINED TO ENSURE LONG TERM SEAL INTEGRITY.

4.0 FOLLOWUP ACTIOPS 4.1 IDENTIFICATION AND EARLY CONTROL 4.1.3 IF HPI HAS INITIATED BECAUSE OF LOU PRESSURE, CONTROL HPI IN ACCORDANCE WIT 11 STEP 3.3.

4.1.2 IF BOTH HPI TRAINS llAVE NOT ACTUATED ON ESFAS SIGNAL, SIART SECOND HPI TRAIN IF POSSIBLE.

BALANCE HPI FLOWS.

4.1.3 IF RC PRESSURE DECREASES CONTINUOUSLY, VERIFY THAT CORE FLOOD TANK (CFTs) AND LOW PRESSURE INJECTION (LPI) HAVE ACTUATED AS NEEDED, AND BALANCE LPI.

o 4.1.4 IF CAUSE FOR C00LDOWN/DEPPISSURIZATION IS DETERMINED TO BE DUE TO A NON-LOCA OVERC00 LING EVDIT AND THE RCS IS AT LEAST 50 F SUBC00 LED THEN PROCEED TO SECTION 4.4.

4.1.5 ATTEMPT TO LOCATE AND ISOLATE LEAK IF POSSIBLE.

LETDOUN WAS ISOLATED IN STEP 2.2 OTHER ISOLATAELE LEARS API PORV (CLOSE BLOCK VALVE) AND BETWEEN VALVES IN SPRAY LINE (CLOSE SPRAY AND BLOCK VALVE).

4.1.6 DETERMINE AVAILA3ILITY OF REACTOR CCOLANT PUMFS (RCPs) A O MAIN AND AUXILIARY FEEDWATER SYSTEMS.

IF FEEDWATER IS NOT AVAILABLE GO TO 4.2.

IF FEED 07ATER IS AVAILABLE GO TO 4.3.

4.2 ACTIONS IF FEEDWATER IS NOT AVAILABLE 4.2.1 THROUGHOUT THE FOLLOWING STEPS MAIT:'AIN MAXDfUM HFI FLOW AND RESTORE FEEDWATER AS SCON AS POSSIBLE.

o 4.2.2 IF RCPs ARE OPERATING, GO TO CNE PUMP PER LOOP.

IF RCPs ARE NOT OPERATING, CO TO STEP 4.2.6 BELOW.

4.2.3 IF RCS PRESSURE IUC'1 EASES, OPEN FORV AND LEAVE OPEN.

NOTE:

IF THE PORV CA!!NOT BE ACTUATED, THE SAFETIES WILL RELIEVE PRESJURE.

1473 021

-s-

69-1106001 o

4.2.4 WHEN FEEDWATER IS RECOVERED, RESTORE OTSG LEVELS IN A CON-TROLLED MANNER.

CLOSE PORV OR BLOCK VALVE, IF POSSIBLE, AND PROCEED TO STEP 4.3.2.

4.?.5 IF NO RCPs ARE OPERATING, OPEN FORV AND MAINTAIN HPI FLOW.

NOTE: IF THE PORV CANNOT BE ACTUATED, THE SAFETIES WILL RELIEVE PRESSURE.

4.2.6 WHEN FEEDWATER FLOW IS RESTORED, RAISE OTSG LEVELS TO 95%

ON THE OPERATE RANGE, CLOSE FORV OR BLOCK VALVE, IF POSSIBLE.

NOTE: OTSG LEVEL SHOULD BE MONITORED PERIODICALLY DURING THE FILL PROCESS.

LEVELS > 95% ON THE OPERATING RA'GE MUST BE AVOIDED TO PRECLUDE FEEDIATER CARRYOVER TO THE STEAMLINES.

4.2.7 VERITY NATURAL CIRCULATION IN THE RCS BY OBSERVING:

o 4.2.7.1 COLD LEG TE:@ERATURE IS SATURATION TEMPERATURE OF SECONDARY SIDE PRESSURE WITHIN APPROXD!ATELY 5 MINUTES.

4.2.7.2 PRIMARY b.T (THOT - TCOLD) BEC0!ES CONSTANT 4.2.8 GO TO STEP 4.3.4.1.

4.3 ACTIONS WITH FEELWATER AVAILABLE TO ONE OR BOTH GENERATORS 4.3.1 MAINTAIN ONE RCP RUNNING PEE LOOP (STOP OTHER RC?s).

IF NO RCPs OPERATING (DUE TO A LOSS OF OFFSITE POWER OR DUE TO MANUAL SECUREMENT PER SECTION 2.0), GO TO STEP 4.3.4 BELOW.

4.3.2 ALLOW RCS PRESSURE TO STABILIZE.

4.3.3 ESTABLISH AND MAINTAIN OTSG C00LI!iG BY ADJUSTING STEAM PRESSURE VIA TURBINE BYPASS AND/0R ATMOSPHERIC DU:IPS.

C00MOU:: AT 100 F PER HOUR TO ACHIEVE AN RC PRESSURE OF 250 PSIG.

REFER TO PRE-CAUTION 3.10 FOR DEVELOP >CNT OF TE:GERATURE AND PRESSURE PLOTS.

ISOLATE CORE FLOOD TANKS tiHEN 50 F SUBC00 LING IS ATTAINED AND RC PRESSURE IS LESS GIAN 700 PSIG.

GO INTO LPI COOLING PER

~

APPENDIX A.

o 4.3.4 IF RCPs ARE NOT OPERATING:

4.3.4.1 ESTABLISH AND CONTROL OTSG LEVEL TO 95% ON THE OPERATE RANGE. VERIFY THE CONDITIONS IN STEP 4.2.7 NOTE: OTSG LEVELS GREATER '"iL\\'i 95% ON THE OPERATI::G RANGE MUST BE AVOIDED TO PRECLUDE FEEDUATER CARRYOVER INTO THE STEAMLINES.

Revised 1473 022 69-1106001 4.3.4.2 IF RC PRESSURE IS DECREASING, WAIT UNTIL IT STABI:.IZES OR BEGINS INCREASINC.

'.E IT BEGINS INCREASING, CD TO STEP 4.3.4.4 4.3.4.3 PROCEED WITH A CONTROLLED C00LDOWN AT 100 F/UR BY CONTROLLING STEAM GENERATOR SECONDARY SIDE PRESSURE.

MONITOR RC PRESSURES AND TCIPERATURES DURING CCOLDOWN AND PROCEED AS INDICATED BELOW:

4.3.4.3.1 IF RC PRESSURE CONTINUES TO DECREASE, FOLLOWING SECONDARY SIDE PRESSURE DECREASES AND WITH PRDIARY SYSTDi TDTERATURES INDICATING SATURATED CONDITIONS, CONTINUE C00LDOWN UNTIL AN RC PRISSURE OF 150 PSI IS REACHED, AND PROCEED TO STEP A.4 0F APPENDIX A.

4.3.4.3.2 IF RC PRESSURE STOPS DECREASING IN RESPONSE TO SECONDARY SIDE PRESSURE DECREASE K;D REACTOR SYSTDI BECOMES SUBC00I.ED, CHECK TO SEL THAT THE FOLLOWING CONDITIONS ARE BOTH SATISFIED:

A) ALL HOT AND COLD LEG TDGERATURES ARE BELOW THE SATURATION TDTERATURE FOR THE EXISTING RCS PRESSURE.

AND B)

THE HOT AND COLD LEG TDTERATURES ARE DECREASING IN RESPONSE TO STEAM GENERATOR SECONDARY TDTERATURE DECREASE.

IF THESE CONDITIONS ARE SATISFIED MiD REMA2; SATISFIED, CONTINUE C00LDOWN TO ACHIEVE AN RCS TDIPERATURE (COLD LEG) 0F 280 F, MiD PROCEED TO STEP A.1 0F APPENDIX A.

NOTE:

IF THE CONDITIONS AE0VE ARE MIT 3ELOW 700 PSIG, THE CORE FLOOD TN;KS SHOULD C

BE ISOLATED.

NOTE:

IF THE PRULGY SY3TDi IS 50 F SUEC00L"D IN bOTH HOT MID COLD LEGS AND PRD'ARY 7

1473 023

69-1106001 SYSTDI PRESSURE IS AB0t.

30 PS'.G, STARTING A REACTOR C00LMT PUST IS PER-IIISSIBLE.

IF SYSTDI DOES NOT RETURN TO AT LEAST 50 F SUSC00 LING IN D70 MINUTES, TRIP PUITS.

IF FORCED CIRCULATION IS ACIIIEVED, PROCEED TO STEP 4.3.

I 1473 024

69-1106001 4.3.4.3.3 IF RC PRESSURE STC. 5 DECREASING AND THE CONDITIONS OF 4.3.4.3.2 ARE NOT MET OR CEASE TO BE MET OR IF RC PRESSURE BEGINS TO INCREASE, THEN PROCEED TO STEP 4.3.4.4 BELOW.

4.3.4.4 RESTORE RCP FLOW (ONE PER LOOP) WHEN POSSIELE PER THE INSTRUCTIONS BELOW.

IF RC PDfPS CANNOT BE OPERATED AND PRESSURE IS INCREASING, GO TO STEP 4.3.4.6.

4.3.4.4.1 IF PRESSURE IS INCREASING, STARTING A PUMP IS PERMISSIBLE AT RC PRESSURE GREATER ~HAN 1600 PSIG.

4.3.4.4.2 IF REACTOR COOLANT SYSTEM PRESSUP2 EXCEEDS STEAM GENERATOR SECONDARY PRESSURE 3Y 600 PSIG OR MORE "BCT" ONE REACTOR C00LANI PUMP FOR A PERIOD OF APPROXHIATELY 10 SECONDS (PREFERABLY IN OPERABLE STEAM GENERATOR LOOP).

ALLOW REACTOR CCOLANT SYSTEM PRESSURE TOSTABILIZE.

CONTINUE C00LDOWN.

IF REACTOR COOLANT SYSTEM PRESSURE AGAIN EXCEEDS SECONDARY PRESSURE DY 600 PSI, WAIT AT LEAST 15 MINUTES AND REPEAT THE PUMP " BUMP".

BUSE ALTERNATE PUMPS SO TEAT NO PUMP IS 3 DEED MORE THAN ONCE IN M HCUR.

THIS MAY BE REPEATED, WITH AN INTERVAL OF 15 MINUTES, UP TO 5 TDfES. AF m THE FIF~H

" BUMP," ALLOW THE REACTOR C00LMT PDT TO CONTINUE IN OPERATICN.

4.3.4.4.3 IF PRESSURE HAS STASILIZED FOR GREATEL TILW ONE HOUR, SECONDARY PRESSUPI IS LESS THAN 100 PSIG AND PRDIARY PRESSURE IS GREATER THAN 250 PSIG, 3USG A PUMP, WAIT 30 MINUTES, AND START AN ALTERNATE PQ2. }

69-1106001 4.3.4.5 IF FORCED FLOW IS ESTABLI511ED, GO TO STEP 4.3.3.

4.3.4.6 IF A REACTOR COOLANT PUMP CANNOT BE OPERATED AND REACTOR COOLANT SYSfEM PRESSURE REACHES 2300 PSIG, OPEN PRESSURIZER PORV TO REDUCE REACTOR COOLANT SYSTEM PRESSURE. RECLOSE PORV WHEN RCS PRESSURE FALLS TO 100 PSI ABOVE THE SECONDARY PRESSURE.

REPEAT IF NECESSARY.

IF PORY IS NOT OPERABLE, PRESSURIZER SAFETY VALVES WILL RELIEVE OVERPRESSURE.

4.3.4.7 MAINTAIN RC PRESSURE AS INDICATED IN 4.3.4.6 IF PRESSURE INCREASES. MAINTAIN THIS COOLING MODE UNTIL AN RC PUMP IS STARTED OR STEAM GENau TOR COOLING IS ESTABLISHED AS INDICATED BY ESTABLISHING CONDITIONS DESCRIBED IN 4.3.4.3.1 OR 4.3.4.3.2.

WHEN THIS OCCURS, PROCEED AS DIRECTED IN THOSE STEPS.

GO TO STEP 4.3.2 IF FORCED FLOW IS ESTABLISHED.

1473 026 e

4 69-1106001 4.4 _NON-LOCA OVERc00 LING TRANSIENT WITH FEEDUATER AVAILABLE 4.4.1 DDlEDIATELY RESTART A RC PUMP IN EACH LOOP IF THE RCS IS 50 F SUEC00 LED.

4.4.2 CONTROL STEAM PRESSURE VIA TURBINE BYPASS OR An!0 SPHERIC DUMP VALVES TO STABILIEE OR CONTROL PLANT HEATUP.

NOTE: CONSIDERABLE HPI MAY HAVE BEEN ADDED TO THE RCS.

THEREFORE, TO PREVENT RCS FROM COING SOLID, THE ABOVE ACTION MAY BE NECESSARY.

4.4.3 AS LONG AS THE RCS IS MAU'TAINED 50 F SUBC00 LED, THROTTLE HPI/MU AND LETDOWN FLOW TO MAINTAIN PRESSURIEER LEVEL AT % 100 INCHES.

4.4.4 USING TURBINE BYPASS VALVES AND FEEDUATER SYSTDf, CONTROL STL1M CENERATORS AS NEEDED TO LDIIT PLANT HEATUP UNTIL RC PRESSURE CONTROL CAN BE RE-ESTABLISHE3 WITH THE PRESSURIEER.

NOTE:

COLD RCS WATER MAY HAVE BEEN ADDED TO THE PRESSURIEER; THEREFORE, A PERIOD OF TDIE MAY ELAPSE BEFORE NCRMAL RC PRESSURE CONTROL CAN BE ESTABLISHED WITH THE PRESSURIEER HEATERS.

4.4.5 ONCE PRESSURE CONTROL IS RE-ESTABLISHED, USE NORMAL HEATLT/

C00LDOWN PROCEDURE TO ESTABLISH DFSIRED PLANT CONDITIONS.

1473 027 :. :.

..,.. -. - _.n -. -,, -

69-1106001 4.5 ACTIONS FOR INADEQUATE CORE __ COOLING 4.5.1 IMMEDIATE STEPS FOR INADEQUATE CORE COOLING NOTE:

IF RC PUMPS ARE RUNNING, DO NOT TRIP PUMPS.

THIS SUPERCEDES INSTRUCTIONS IN SECTION 2.1.

4.5.1.1 VERIFY HPI/LPI SYSTEMS ARE FUNCTIONING PROPERLY WITH MAXIMUM FLOW.

START MAKEUP PUMP (S), IF POSSIBLE, TO INCREASE INJECL' ION FLOW.

4.5.1.2 VERIFY STEAM GENERATOR LEVEL IS BEING CONTROLLED

~

AT 95% ON OPERATE RANGE.

NOTE:

FOR TECO STEAM GENERATOR LEVEL SHOULD BE AT 96 INCHES INDICATED ON THE STARTUP RANGE CAUTION:

REFERENCE LEG BOILING COULD GIVE

.,.h FALSE LEVEL INDICATION 4.5.1.3 DEPRESSiJRIZE OPERATIVE STEAM GENERATOR (S) TO ESTABLISH A 1000F/HR DECREASE IN SECONDARY Z-il SATURATION TEMPERATURE.

p 4.5.1.4 ENSURE CORE FLOOD TANK ISOLATION VALVES ARE OPEN.

4 4.5.1.5 IF REACTOR COOLANT SYSTEM PRESSURE INCREASES TO 2300 PSIG (1500 PSIG FOR DB-1) OPEN PRESSURIZER y

PORV TO REDUCE REACTOR COOLANT SYSTEM PRESSURE.

RECLOSE PORV WHEN RCS FALLS I] 100 PSIG ABOVE i

THE SECONDARY PRESSURE.

REPEAT IF NECESSARY.

43' IF PORV IS NOT OPERABLE, PRESSURIZEP. SAFETY 9'j VALVES WILL RELIEVE PRESSURE.

4.5.1.6 PROCEED IMMEDIATELY TO 4.5.2.

4.5.2 WHEN THE INDICATED INCORE THERM 0COU?LE TEMPERATURES OR 1

HOT LEG RTD TEMPERATURES ARE SUPERHEATED FOR THE EXI; TING RCS PRESSURE, OPERATOR ACTION SHALL BE BASED ON CONDITIONS DETERMINED FROM FIGURE 3, BY A SAMPLE OF THE HIGHEST INCORE THERMOCOUPLE TEMPERATURE READINGS TO DETERMINE THE CORE EXIT THERMOCOUPLE TEMPERATURE.

O!

NOTE:

MORE THAN ONE THERM 0COGPLE TEMPERATURE READI"G

!t SHOULD BE USED (FOR EXAMPLE USE THE AVERAGE OF 5).

63-1106001 4.5.3 WHEN THE INCORE THERMOCOUPLE TEMPERATURE HAS BEEN DETERMINED PER SECTION 4.5.2, GO TO THE SECTION INDICATED BELOW.

INCORE THERMOCOUPLE TEMPERATURE SECTION l

INCORE Tc 1 SATURATION 4.1.6 j *F CURVE 1 A.INCORE Tc < CURVE 2 FIGURE 3 4.5.4 2

+

INCORE Tc R CURVE 2 FIGURE 3 4.5.5 l

NOTE:

THE INCORE THERMOCOUPLE TEMPERATURE READINGS

.a SHALL BE CONTINUOUSLY MONITORED UNTIL THE INDICATED INCORE THERMOCOUPLE TEMPERATURES RETURN TO SATURATION TEMPERATURE FOR THE EXISTING RCS S,rt PRESSURE.

4.5.4 ACTIONS FOR CURVE 11 INCORE Tc < CURVE 2 FIGURE 3 4.5.4.1 IF RC PUMPS ARE NOT OPERATING, START ONE PUMP PER LOOP (IF P9SSIBLE).

THIS INSTRUCTION SUPERSEDES PREVIOUS INSTRUCTIONS TO TRIP d

RC PUMPS.

T*E NOTE:

DO NOT BYPASS NORMAL INTERLOCKS.

4.5.4.2 DEPRESSURIZE OP.ERATIVE STEAM GENERATOR (S) AS 4.

RAPIDLY AS POSSIBLE TO 400 PSIG OR AS FAR AS NECESSARY TO ACHIEVE A 1000F DECREASE IN SECONDARY SATURATION TEMPERATURE.

4.5.4.3 OPEN THE PORV, AS NECESSARY, TO MAINTAIN RCS 7.j PRESSURE WITHIN 50 PSI 0F STEAM GENERATOR

f SECONDARY SIDE PRESSURE.

NOTE:

IF STEAM GENERATOR DEPRESSURIZATION a

WAS NOT POSSIBLE, OPEN PORV AND LEAVE OPEN.

4.5.4.4 IMMEDIATELY CONTINUE PLANT C00LDOWN BY MAIN-TAINING 100F/HR.

DECREASE IN SECONDARY

{

SATURATION TEMPERATURE TO ACHIEVE 150 PSIG j

RCS PRESSURE.

CAUTION:

IF AUXILIARY Fetu PUMP IS SUPPLIED 3Y MAIN STEAM, DO NOT DECREASE PRESSURE BELOW THAT PRESSURE NECESSARY

.h 1

{,}1 FOR AUXILIARY FEED PUMP OPERATION.

3 473 OM 69-1106001 4.5.4.5 IF THE AVERAGE INCORE THERMOCOUPLE TEMPERATURE i

INCREASES TO CURVE 2 FIGURE 3 PROCEED IMMEDIATELY TO SECTION 4.5.5.

y 4.5.4.6 WHEN RCS PRESSURE REACHES 150 PSIG, GO TO APPENDIX "A".

4.5.5 ACTIONS FOR I CORE Tc 1 CURVE 2 FIGURE 3 4.5.5.1 IF POSSIBLE, START ALL RC PUMPS.

NOTE:

STARTING INTERLOCKS SHOULD BE DEFEATED g

IF NECESSARY.

[y 4.5.5.2 DEPRESSURIZE THE OPERATIVE STEAM GENERATOR (S)

AS QUICKLY AS POSSIBLE TO ATMOSPHERIC PRESSURE.

CAUTION:

IF AUXILIARY FEED PUMP IS SUPPLIED BY MAIN STEAM, DO NOT DECREASE i

PRESSURE BELOW THAT PRESSURE NECESSARY FOR AUXILIARY FEED PUMP OPERATION.

4.5.5.3 OPEN THE PRESSURIZER PORV AND LEAVE OPEN.

[

NOTE:

THE RCS WILL DEPRESSURIZE AND THE LPI SYSTEM SHOULD RESTORE CORE COOLING 4.5.5.4

.WHEN INCORE THERMOCOUPLE TEMPERATURES RETURN

'r.

TO THE SATURATION TEMPERATURE FOR THE EXISTING

[

RCS PRESSURE - AND - THE LPI SYSTEM IS DELIVERING FLOW, PROCEED AS FOLLOWS:

y f

4.5.5.4.1 CLOSE THE PRESSURIZER PORV; REOPEN IF RCS PRESSURE INCREASES ABOVE w

150 PSIG.

4.5.5.4.2 DECREASE TO TWO (2) RC PUMP OPERATION j

(ONE PER LOOP).

F 4.5.5.4.3 ISOLATE THE CORE FLOOD TANKS.

4.5.5.4.4 MAINTAIN STEAM GENERATOR PRESSURE AT 4

ATMOSPHERIC OR AS LOW AS POSSIBLE IF Fr MAINTAINING AUXILIARY FEED PUMP IN OPERATION OFF OF MAIN STEAM.

7 4.5.5.4.5 CONTROL HPI PER 3.3.

55 473 030 69-1106001 AR P ROACH L

N SYSTEM FOR SUCTION FROM RB SUMP.

CLOSE THE LPI BWST SUCTION VALVFS.

h NOTE:

IF HPI IS REQUIRED PER 3.3, i

ALIGN LPI AND HPI IN PIGGYBACK MODE.

CLOSE HPI SUCTION VALVES i

TO BWST.

4.5.5.4.7 GO TO APPENDIX "A".

e 1.473 031 69-1106001 APPENDIX A

,LPI COOT,I:!G A.1 DETER:!INE IF PRIMARY COOLANT IS AT LEAST 50 F SUECCOLED.

IF NOT, GO TO STEP A.3.

~

A.l.1 START LPI PUMPS. IF E0TIl PETS AP2 OPERADLE, GO TO STEP A.2.

FOR ONE LPI PMIP OPERABLE MAINTAIN OTSG CCOLI:'G M;D PRCCEED AS FOLL01.'S.

THE OPERABLE LPI PU!T WILL BE USED TO l'AE TAI:l SYSTC: n:VE::10RY.

A.1.2 OBTAIN PRDIARY SYSTEM CONDITIO"3 0F s 280F AND s 250 PSIG.

g A.l.3 ALIGN THE DISCIIARGE OF THE OPERAELE LPI PCT TO THE SUCTICNS OF THE HPI PDTS AND TAKE SUCTION FROM THE BWST.

IF TIIE BWST IS AT THE LOU LEVEL ALARM, ALIGN LPI SUCTION FROM THE RB SMT N D SiiUT SUCTION FROM BWST.

A.l.4 START Tile OPERAELE LPI PET SPECIFIED AE0VE.

THE H?I-LPI SYS*; DIS UILL NOW BE IN " PIGGY BACK" AND 11PI FLOU IS MAINTAINING SYSTC!

PPISSURE.

A.l.5 GO TO SINGLE RC PD:P OPERATION.

A.l.6 Ul!EN THE SECOND LPI PUMP IS AVAILABLE, ALIGN IT IN Ti1E DECAY HEAT If0DE AND COMME;CE DECAY HEAT'RDIOVAL.

(DECAY HEAT SYSICI FLOW GREATER THAN 1000 GPM).

SECURE RE!AE U:G RC PCIP UHEN DECA'l HE'.T RDIOVAL IS ESTA3LISFID.

CAUTION: VERIFY THAT ADEQUATE NPSH.NISTS FOR THE DECAY HEAT PUMP IN THE DH RE! OVAL MODE.

IF INADEQUATE, TRA'SFER TO LPI MODE.

A.l.7 REDUCE REACTOR CCOLANT FRESSURE TO 150 PSIG BY THROTTLI::G HPI FLO',!.

CONTROL RC TCTERATURE USE G THE DECAY !! EAT SYSTCf CCOLER BYPASS TO IIAINTAIN SYSTCt PRESSURE AT LEAST 50 PSI ABOVE SATU"ATION PRESSURE, TO ASSUP.E THAT :PSU REQUIRC:1.::TS FOR TIIE DECAY HEAT PU'IP ARE MAINTAINED.

om

• j' 9' Q' g g

D oJu_2.l\\.Uh o o Ju

69-1106001 A.1.8 SI: CURE Tile !!PI PUMP M;D Sli1FT Tl!E LPI Pun' SUPPLY 1:;0 LT TO Tile LPI INJECTION 1:0DE.

j A.1.9 REDUCE REACTOR COOL &'T TDTERATURE TO 100 F BY CO:;TROLLI :G Tl!E DECAY llEAT SYSTDI COOLER BYPASS.

1:0TE:

IF O!'E OF T11E LPI/ DECAY llEAT PC:PS IS LOST, RETUPJ: TO OTSG COOLING USING NATURAL CIRCULATION OR ONE REACTOR CCOLA::T PUMP (A1).

A.2 COOLDOWN ON TWO LPI PUMPS A.2.1 MAINTAIN RCS PRESSURE AT s 250 PSIG AND REDUCE RCS TEMPERATURE T0 s 280F.

A.2.2 ALIGN ONE LPI PLTIP IN Tile DECAY HEAT REIOVAL MODE.

A.2.3 SECUP2 CNE RC PUMP IF TWO ARE OPERATT.::G.

A.2.4 START THE DECAY !! EAT PLTT IN THE DECAY HEAT RCIOVAL MODE, A';D WHEN DECAY HEAT SYSTE! FLOW IS GREATER THA'? 1000 'GPM, SECURE THE RC,7;I::G RC PUMP.

A.2.5 REDUCE RC PPISSUP2 TO 150 PSIG BY THROTTLI::G HPI FLOW.

CO:: TROL RC TDIPERATURE TO MAINTAIN AT LEAST 50 PSI MARGI'i TO SATURATIO:: PP2SSUP2.

A.2.6 START THE SECOND LPI PUMP IN THE LPI n!JECTIO:: MODE.

SECURE UPI PC2.

A.2.7 SHIFT LPI SUCTIO:; FROM THE EWST TO THE REACTOR BUILDING SC 2 kiIEN SUFFICIC'T NPSH IS AVAILABLE.

NOTE:

TilIS IS DESIRABL2 TO AVOID U::::ECESSARY QUANTITIES OF WATER IN CONTAINMENT.

A.2.8 REDUCE REACTOR COOLANT TD!?ERATURE TO 100 F BY CONTROLLn;G IHE DECAY llEAT SYSTDi COOLER SYPASS.

NOTE:

IF ONE OF THE LPI/ DECAY llEAT PtT1PS IS LOST, RETUPJ: TO OTSG COOLING USING NATURAL CIRCULATION OR 0 ;E RC PC9 PER A.1.

1 - 1473 033

69-1106001 A.3 COOL DOU:: l'C S'.'STC AT StTURATIO::'

l A.3.1 MAINTAIN RC PRESSURE AT s 250 PSIG.

E A.3.2 ALIGN ONE LPI PUMP EO SUCTION Of Tlin 11PI PU"PS A :D Tlin SUCTIO : IO TI!E REACTOR BUILDI::G SU:!P.

(S11UT Ts:ST SUCTIO:! VALVE FCR TilIS PCIP.)

A.3.3 WilEN Tile BUST LEVEL REACHES T!IE LO-LO LEVEL LD:ITS, START Tile LPI PUMP AND SilUT Ti1E IIPI PL71P SUCTION FRC:: Ti!E ENST.

A.3.4 WilEN PRI!!ARY SYSTDI TD!PERATURE BECOMES SUDC00 LED BY AT LEAST 50 F, CO TO A.1.1.

A.4 COOLDOUN WIT!!OUT REACTOR COOLANT PU:TS A.4.1 RCS INITIAI. CONDITIO :S ARE:

PRESSURE 150 PSI, TE:PERATURE AT SATURATION.

A.4.2 ALIGN LOW PRESSURE INJECTION SYSTDI FOR SUCTION FROM REACTOR SUILDI::G SUMP AND PLACE INTO SERVICE.

A.4.3 BALANCE LPI INJECTION AND CO:: TROL RC TE:IPERATURE WITH DECAY HEAT COOLERS.

A.4.4 ISOLATE CORE FLOOD TA'fKS.

A.4.5 GO TO STEP A.l.1 AND FOLLOW TIIE PROCEDURE GIVEN TiiERE, IG: 0 RING Tile INSTRUCTIONS RELATI:G TO RC PCIP OPERATION.

Of D

DIS [h

%o e) $ b.\\ lTU a 1473 034

_,e_

e*. %

... %== w w _ __

_=e_

_____es===

Figure 1 Pressure-Temperature Limit Curve to Preclude Reactor icssel Britric Fracture during RCS Depressurization Follouing Accident Conditions.

Applicability:

2 EFPY of Operation bcyond 6/79 T

l I

t

+l..

l l

(<

')

2500 a

r n

El

.i j.

Note:

g Adjustments for possible instrumenta-7

- j --

g tion error and elevation Pressure l

..-g.

differentials have been incorporated

-l-l i

j q

,)

into this P-T Limit Curve o

._..u.

..s.

.i y

.e 00 i

a

. =. =.

.r

..n.

c>.

o

t..

s.

e y,

.o u

t1 u

s.

c.

t2 c

n 1500

_.f.

c

......I

.l

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_.g.._

_.._... _... _.3 U

o

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. ~. _..

m 5.,.

g s

em res3e e

u l,

.r q

s r2 u

.t..

g

.g..

0 50 I

U

\\

a 1000 s

..s l.

.o

.t.

.t.

O, c3

.l.

a naCCc Es.

c o

a,

.s_

o

~

210 2214 m.

218 2500 i

.l..

y r.*

w:

_t_

o.

^

l

-l 1

500 m-

. L..

~

t.

~

o.

I pa

.._.l..__

.. T..

_ L.:

Acceptable Ss

.t

.t

.i l.

c3 0

~

U 60 80-100 120 140 160 180

.'00 220 tn Reactor Vessel Dwncomer Temperature, 'F

69-1106001 FIGURE 2 FOR TECO ONLY NOT NEEDED FOR LOWERED LOOP PLANTS 1473 036 69-1106001 Figure 3 CORE EXIT THERMOCOUPLE TE.'aPERATURE FOR INADEQUATE CORE COOLING 1200 o'

1100 CURVE #2 5

i

{

TCLAG LESS THAN 1600*F f

1000 8

E 900 lE M"

800 CllRVE #1 r

O u

TCLA0 LESS THAN 1400 F 700 t

600 500 400 200 600 1000 1400 1800 2200 Pressure, psia 1473 037 5

69-1106001 Part II:

Small Break phenomena - Description of plant Behavior 1.

Introduction A los.-of-coolant

.ent is a condition in which liquid inventory is lost from the a_. or coCsnt system.

Due to the loss of mass from the reactor coolant system. *'

cost sipnificant short-tem symptom of a loss-of-coolant acefdeat is an uncontrolled reduction in the reactor coolant system pressure.

The reactor protection system is designed to trip the :eactor on low pressure.

This should occur before the reactor coolant system reaches saturation conditions.

The existence of saturated conditions within the reactor system is the principal longer-term indication of a LOCA and requires special consideration in the development of operating procedures.

Following a reactor trip, it is necessary to remove decay heat from the reactor' core to prevent damage. However, so long as the reactor core is kept covered with cooling water, core damage will be avoided. The ECCS sNtems are designed to respond automatically to low reactor coolant pressure conditions and take the inital actions to protect the reactor core. They are sized to provide sufficient water to keep the reactor core covered even with a single failure in the ECCS systems.

Subsequent operator actions are required ultimately to place the plant in a long-term cooling mode. The overall objective of the automatic emergency core cooling system and the followup operator actions is to keep the reactor core cool.

A detailed discussion of the small break LOCA phenomenalogy is presented in this section.

This discussion represents Part II of the operating procedure guidelines for the development of detailed operating procedures.

part I presents the more detailed step-by-step guidelines.

i473 038

-22_

1106001 The response of the primary system to a small break will greatly depend on break size, its location in the system, operation of the reactor coolant pumps, the number of ECCS trains functioning, and the availability of secondary side cooling.

RCS pressure and pressurizer level histories for various combinations of parameters are presented in order to indicate the wide range of system behavior which can occur for small LOCA's.

Impact of RC pump Operation on a Small LOCA 2.

With the RC pumps operating during a small break, the steam and water will remain mixed during the transient. This will result in liquid being discharged out the break continuously. Thus, the fluid in the RCS can evolve to a high void fraction.as shown in Figure 1.

The maximum void fraction that the system evolves to, and the time it occurs, is dependent on the break size and locatior. Continued R6 pump ope:ation, even at high system void fractions, will provide sufficient core flow to keep cladding temperatures within a few degrees of the saturated fluid temperature.

, Since the RC$ can evolve to a high void fraction for certain small breaks with the RC pumps on, a RC pump trip by any means (i.e., loss of offsite power, equipment failure, etc.) at a high void fraction during the small break transient may lead to inadequate core cooling. That is, if the RC pumps trip at a time period when the system void fraction is greater than approximately 70%, a core heatup will occur because the amount of water left in the RCS would not be sufficient to keep the core covered.

The cladding temperature would increase until core cooling is re-established by the ECC systems.

For certain break sizes and times of RC pump trip, acceptable peak cladding temperatures during the event could not be assured and the core could be damaged.

Thus, promot operator action to trip the RC pumps upon receipt of a low pressure ESFAS signal is required in order 14~/ 3 039 69-1106001 to ensure that adequate core cooling is provided.

Following the RC pump trip, the small break transient will evolve as described in the subsequent sections.

3.

Small Breaks with Auxiliary Feedwater There are four basic classes of 3reak response for small breaks with auxiliary feedwater.

These are:

~

1.

LOCA largd enough to depressurize the reactor coolant system 2.

LOCA which stabilizes at approximately secondary side cressure 3.

LOCA which may repressurize in a saturated condition 4.

Small LOCA which stabilizes at a primary system greater than secondary system pressure The system transients for these breaks are depicted in Figure.2.

3.1 LOCA large Enough to Depressurize Reactor Coolant System,: Cumes 1 aiid 2 of Figure 2 show the response of RCS pressure to breaks that are large enough in combination with the ECCS to depressurize the system to a stable low pressure.

ECCS injection easily exceeds core boil-off and ensures core cooling. Curves 1 and 2 of Figure 3 show the pressurizer level

~

transient.

Rapidly falling pressure causes the hot legs to saturate quickly. Cold leg temperature reaches saturation somewhat later as RC pumps coast down or the RCS depressurizes below the secondary side saturation pressure.

Since these breaks are capable of depressurizing the RCS without aid of the steam generators, they are essentially unaffected by the availability of auxiliary feedwater. U;,un receipt of a low pressure ESFAS signal, the operator must trip all RC pumps and verify that all ESFAS actions have been co.mpleted. The operator must also balance HPI flows such that flow is available through all HPI injection nozzles even if only one HP1 is available.

The operator should also ba. lance LPI flows, shouTd the system be actuated, to ensure flow through both lines. The operator needs to take no further actions to bring the system to a safe shutdown 1473 040 69-1106001 condition. Rapid depressurization of the steam generetors would only act to accelerate RCS depressurization.

It is, however, not necessary.

Restarting of the RC pumps is not desirable for this class of break.

Long-term cooling will require the operator to shift the LPI pumo suction to the reactor building sump.

3.2 LOCA Which Stabilizes at Approximately Secondary Side Pressure. Curve 3 of Figure ? shows the pressure transient for a break which is too small in combination with the operating HPI to depressurize the RCS. The steam generators are, therefore, necessary to remove a portion of core decay heat. Although the systcm pressure will initially stabilize near the secondary side pressure, RCS pressure may eventually begin falling as the decay heat level decreases. Curve 3 of Figure 3 shows pressurizer level behavior. The hot leg. temperature quickly equalizes to the saturated temperature of the secondary side and controls primary system pressure at saturation. The cold leg temperature may remain slightly subcooled.

If the HPI refills and repressurizes the RCS, the hot legs can become subcooled. The immediate operator action is to trip the RC pumps upon receipt of the low pressure ESFAS signal and then verify ESFAS functions.

The. operator must ther balance HPI in order to ensure flow through each high pressure injection line.

Followup action by the operator is to raise the emergency feedwater level to 95% on the operating range and check for established natural circulation. This is done by gradually depressurizing the s m

generaton.

If this test fails, intermittent bumping of a RC pump should be performed as soca as one is available. Contir.ued depressurization of the steam generators with nat tral circulation leads to cooling and depressurization of the RCS. The operator's goal is to depressurize the RCS to a pressure that enables the ECCS to exceed core boil-off, possibly refill the RCS, and to ultimately establish long-term cooling.

1473 041 69-11C6001 3.3 LOCA Which May Repressurize in a Saturated Condition.

Curve 4 of Figure 2 shows the behavior of a small break that is 400 small, in combination with the HPI, to depressurize the primary system. Although steam generator feedwater is available, the loss of primary system coolant and the resultant RCS voiding will eventually lead to interruption of n'atural circulation. This is followed by gradual repressurization of the primary system.

It is possible that the primary ~ system could repressurize as high as the pressurizer safety valve setpoint before the pressure stabilizes.

This is shown by the dashed line in Curve 4.

Once enough inventory has been lost from the primary system to allow dircet steam condensation in the regions of the steam generators contacting secondary side coolant, the prima,ry system is forced to depressurize to the saturation pressure of the secondary side.

Since the cooling capabilities of the secondary side are needed to continae to remove decay heat, RCS pressure will not fall.below that on the secondary side.

HPI flow is sufficient to replace the invantory lost to boiling in the core, and mndensation in the steam generators removes decay heat energy. The RCS is in a stable thermal condition and it will remain there until the operator takes further action. The pressurizer level response is characterized by Curve 3 of Figure 3 during the depressurization, and Curve 4 of Figure 3 during the temporary repressurization phase. The dashed line indicates the level behavior if pressure is forced up to the pressurizer safety valve setpoint.

During this transient, hot leg temperature will rapidly approach saturation with the initial system depressurization, and it will remain caturated during the rhole transient.

Cold leg temperature will approach saturation as circulation is lost, but may remain slightly subcooled during the repressurization phase of the transient.

Later RCS depressurization could cause the cold leg temperatures to reach saturation.

Subsequent refilling of the primary 1473 042 69-L106001

~.

system by the HPI might cause temporary interruption of steam condensation in the steam genera. tor as the primary side level rises above the secondary side level.

If the depressurization capability of the break and the HPI is insufficient to offset decay heat, the primary system W il once more repressurize. This decreases HPI flow and increases loss through the break until enough RCS coolant is lost to once more allow direct steam condensation in the steam generator. This cyclic behavior will stop once the HPI and break can balance decay heat or the operator takes some action.

The operator's insnediate action is to trip the RC pumps upon receipt of the low pressure ESFAS signal and verify the completion of all ESFAS functions. The operator should then balance HPI flow.

Following that, he should raise the steam generator level to 95% o'f the opera' ting range and check for natural circulation.

If it is positive, he should depressurize the steam generators, cool and depressurize the primary system, and attempt to refill it and establish long-term cooling.

If the system fails to go into natural circulation, he should open the PORV long enough to bring and hold the RCS near the secondary side pressure. Once natural circulation is established or a RC pump can be bumped, he will be able to continue depressurizing the RCS with the steam generators and establish long-term cooling.

3.4 Small LOCA Which Stabilizes at F> Psec. Curve _5 of Figure 2 shows the behavior of the RCS pressure to a break for which high pressure injection is being supplied and exceeds the leak flow before the pressurizer has emptied. The primary system remains subcooled and natural circulation to the steam generator removes core decay heat. The pressurizer never empties and continues to control primary system pressure.

The operator needs to trip the RC pumps and ensure that ESFAS actions have occured. Throttling of HPI 0

is permitted only after RCS subcooling of 50 F has been established, the pressurizer has refilled, and natural or forced circulation has been 1473 043

-n-

69-1106001 verified. A restart of the RC pumps under these conditions is desirable for plant control.

3.5 Small Breaks in Pressurizer. The system pressure transient for a small break in the pressurizer will behave in a manner similar to that previous':y discussed. The initial depressurization, however, will be more rapid as the initial inventory loss is entirely steam.

~

The pressurizer level response for these accidents will initially behave like a very small break without auxiliary feedwater.

The initial rise in pressurizer level shown in Figure 4 will occur due to the pressure reduction in the pressurizer and an insurge of coolant into the pressurizer from the RCS. Once the reactor trips, system contraction causes a decreasing level in the pressurizer.

Flashing will ultimately occur in the hot leg piping and cause an insurge into the pressurizer. This ultimately fills the pressurizer.

For the remainder of the transient, the pressurizer will remain full. Toward the later stages of the trarsient, the pressurizer may contain a two-phase m'ixture and the indicated level will show that the pressurizer is only partially full. Except for. closing the PORV block valve, operator actions ind system response are the same for these breaks as for similar breaks in the loops.

4.

Small Breaks Without Auxiliary Feedwater There are three basic classes of break response for small breaks without auxiliary feedwater.

These are:

1.

Those breaks capable of relieving all decay heat via the break.

2.

Breaks that' relieve decay heat with both the HPI injection and via the break.

3.

Breaks which do not automatically actuate the HPI and result in system repressurization.

The system pressure transients for these breaks are depicted in Figure 5.

1473 044 69-1106001

~

4.1 LOCA's large Enouoh to Depressurize Reactor Coolant System.

For Class 1 (curve 1.of Figure 5), RC system pressure decreases smoothly throughout the transient.

For the larger breaks in this class, CFT actuation and LPI injection will probably occur.

For the smaller breaks of this class only, CFT actuation will occur. Auxiliary feedwater injection is not necessary for the short-tem stabilization of these breaks. The pressurizer level for this transient rapidly falls off scale. Operator action and plant response are similar to those described for this class of breaks with a feedwater supply.

4.2 LOCA's Which Reach a Semi-Stabilized-State.

For Class 2 (Curve 2 of Figure 5) breaks, the RC pressure will rapidly reach the low pressure ESFAS trip signal (about two to three minutes). With the HPI's on, a slow system depressurization will be established coincident with the decrease in core decay heat. No CFT actuation is. expected. Auxiliary feedwater is not necessary for the shert-tem stabilization of these breaks. The pressurizer level for thi's transient rapidly falls off scale.

The operator needs to trip the RC pumps upon the icw pressure ESFAS signal, verify completion of all ESFAS functions, and try to establish secondary side cooling.

Balancing of the HPI must also be perfomed.

If steam generator feedwater cannot be obtained and RCS pressure is increasing, the operator should open the PORY and provide all the HPI and makeup capability possible.

The goal is to depressurize and cool the core with the ECCS, the PORV, and the break.

If secondary side cooling is again established, the operator should verify natural circulation, and if unavailable, bump a RC pump to complete RCS cooldown with the steam generators.

At this point, the PORV can be closed, the system refilled, and long-tem cooling established.

1473 045 69-1106001 4.3 Small LCJA's Which do not Actuate the ESFAS. Automatic ESFAS actuation will not occur for Class 3 (Curve 3 of Figure 5,) breaks.

Once the SG secondary side inventory is boiled off, system repressurization will occur as the break is not capable of removing all the decay heat being generated in the core.

System repressurization to the PORV or the pres-surizer safety valves will occur for smaller breaks in this class.

Tor the "zero" break case, repressurization to the PORV will occur in the first five minutes. Operator action is required within the first 20 minutes to ensure core coverage throughot.t the transient.

For the 177-FA lowered loop plants, this action can be either manual actuaticn t/ the auxiliary feedwater system or the HPI system.

The establishment of auxiliary feedwater will rapidly depressurize the RCS to the ESFAS actuation pressure, and system pressure will stabilize at either the secondary side SG pressure or at a pressure where the HPI equals the leak rate.

Upon receipt of the low pressure ESFAS signal, the operator must trip the RC pumps.

For the Class 3 breaks, pressurizer level response will be as shown in Figure 6.

The minimum refill time for the pressurizer is that for the "zero" break and is shown in Figtre 6.

After initially drawing inventory from the pressurizer, the system repressurization will cause the pressurizer level to increase, possibly to full pressurizer level. Once the operator action to restore auxiliary feedwater has been taken, the system 3473 046 69-1106001 depressurization will result and cause an outsurge from the pressurizer.

Complete loss of pressurizer level may result.

For the smaller breaks in Class 3 which result in a system repressurization fullowing the actuation of the HPI system, pressurizer level will increar.e and then stabilize.

Without auxiliary feedwater, both the hot and cold leg temperaturrs will saturate early in the transient and, for the Class 1 and 2 breaks, will remain sa'turated. For the Class 3 breaks, once auxiliary feedwater is established, the cold leg temperatures will rapidly decrease to approximately the sautration temperature corresponding to the. SG secondary side pressure and will remain there throughout the remainder of the transient. Hot leg temceratures will remain saturated throughout the event.

Tne operator needs to manually initiate all ESFAS actions, balance HPI flow, and attempt to restore secondary side cooling.

In the meantime, he should actuate the makeup pump and open the PORV in order to cool the core and limit the RCS rep'ressurization. Once feedwater is available, he c:n close the PORV and contintie the RCS cooldown and depressurization with the steam generators.

If natural circulation has not been established, he can bunp a RC pump to cause forced circulation. The goal is to depressurize to where the ECCS can refill the RCS and guarantee long-term cooling.

4.4 Small Breaks in pressurizer. See the writeup for small breaks in pressurizar with feedwater.

Small breaks in the pressurizer will differ from those in the loops in the same manner as those previously described in the section addressing small breaks in the pressurizer with auxiliary feed.

5.

Transients with Initial Resconse Similar to a Small Break Soveral transients give initial alarms similar to small breaks.

These transients will be distinguished by additional alarms and indications or subsequent system response.

Overcooling transients such as steam line breaks, increased feedwater 1473 047

69-1106001 flow, and steam generator overfill can cause RCS pressure decreases with low-pressure reactor trip and ESFAS actuation.

But steam line breaks actuate low steam pressure alams for the affected steam generator, and steam generator overfills result in high steam generator level indications.

The overcooling transients will repressurize the primary system because of HPI actuation, and will return to a subcooled condition during repres-surization. The immediate actions for both overcooling and small break transients are the same, including trippir.g of the RC pumps.

The operator will recognize overcooling events during repressurization, if not sooner, and is instructed to throttle HPI and restart the RC pumps, if subcooled conditions are established, by the small break operating instructions.

A loss-of-feedwater transient will result in a high reactor system pressure alam but does not give an ESFAS actuation alam.

A loss ~of integrated control system power transient starts with a high RC pressure trip. After the reactor trip, this becomes an overcooling transient and will give low reactor system pressure and possible ESFAS actuation. Steant generator levels remain high and the system becomes subccoled during repressurization.

Design features of the B&W NSS provide automatic protection during the early part of small break transients, thereby providing adequate time for small breaks to be identified and appropriate action taken to protect the system. The only prompt manual operator action required is to trip the RC pumps once the low pressure ESFAS signal is reached.

6.

Transients that might Initiate a LOCA There are no anticipated transients that might initiate a LOCA since the PORV has been reset to a higher pressure and wiil not actuate during anticipated transients such as loss of main feedwater, turbine trip, or loss of offsite power.

048

69-1106001 However, if the PORV should lift and fail to reset, there are a number of indications which differentiate this transient.(rom the anticipated transients identified above. These include:

o ESFAS actuation o Quench tank pressure / temperature alams o Saturated primary system o Rising pressurizer level These additional signals will identify to the operator that in addition to the anticipated transient, a LOCA has occurred.

In the unlikely event that small breaks other than a malfunctioning FORV occur after a transient, they can be identified by initially decreasing RCS pressure Small and convergence to saturation conditions in the reactor coolant.

break repressurization, if it occurs, will follow saturation conditions.

By remaining aware of whether the reactor coolant remains subcooled or becomes saturated after transients, the operator is able to recognize when a small break has occurred.

7.

HPI Throttling For small LOCA's, the HPI system is needed to provide makeup to the RCS and must rema,in operable unless specific criteria are satisfied.

W The basis for these criteria are described below.

For certain small breaks, system depressurization will result in LPI actuation. Since the LPI is designed to provide injection at a greater capacity than the HPI, temination of the HPI is allowed. However, this action should only be taken if the flow rate through each line is at least The 20-minute 1000 gpm and the situation has been stable for 20 minutes.

tins delay is included to ensure that the system will not repressurize and result in a loss of the LPI fluid.

In the event of a core flooding line break, the LPI fluid entering the broken core flooding line will not reach the vessel.

Thus, in order to ensure that fluid is continually being injected to the RV for all breaks, the LPI must be providing fluid through both lines. The 1000 gpm is equivalent to the flow from 1473 049 69-1106001 two HPI pumps and ensures that upon termination of the HPI pumps, adequate flow is being delivered to the RV.

Throttling or termination of the HPI flow is also allowed if all the following criteria are met:

0 A.

Hot and cold leg temperatures are at least 50 F below the saturation temperatures for the existing RCS pressure.

B. The action is necessary to prevent the indicated pressurizer level from going off-scale high.

Under these conditions, the prima.y system is sclid. - Continued HPI flow at full capaci'.y may result in a solid pressurizer and would result in a lifting of the PORV and/or the pressurizer code safety valves.

This say in turn lead to a LOCA. Thus, HPI flow should be throttled to 0

maintain a stable inventory in the RCS. However, if the 50 F subcooling cannot be maintained, the HPI shall be imediately reactivated.

HPI flows shduld also be thrt,ttled to pmvent violation of the nil ductility temperature (NDT) for the reactor vessel.

This concern can only arise if the fluid temperature 0

within the reactor vessel is at letst 50 F subcooled. A curve of the allowable downcomer temperature for a given RCS pressure is provided within the operating guidelines. TP.e downcemer temperature is datermined by one of two methods:

1.

If one or more RC pumps are operative, the cold leg RTD reading will be essentially the same as the reactor vessel downcomer temperature.

1473 050 2.

Without the RC pumps operating, the cold leg RTD's may not provide 69-1106001 temperature readings indicative of the actual RV downcomer temperature, as a stagnant pool of water may exist at these locations.

The incore thennoccuples will provide the best indicator of the downcomer temperature and should be :Jtilized if no RC pumps are available.

In order to account for heat added to the fluid fran the core,150'F must be subtracted from the incore thennocouple readings to reflect the downcomer temperature.

This method will result in temperatures which will be lower than the expected downcemer temperature. Thus, use of this methodology assumes that NDT will not be a problem.

0 T A73 OSI 6 ~.

69-1106001 BREAK SPECTRUM AVERAGE SYSTEM V010 F'IACTION WITH THE RC PUMPS OPERATIVE AND 2 HPl PUMPS 100 f"*' wen g

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400 000 1200 1600 2000 2400 2800 Time, sac Figure 1 1473 052 69-1106001 PRESSURE VS TIME-SMALL BREAKS WITH AUXILIARY FEEDWATER

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69-1106001 3

PRESSURIZER LEVEL VS TINE.SWALL BREAKS WITH AUXILIARY FEEDWATER 100 g_ _ q l

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-38_

69J1106001 PPESSURIZER LEVE'. VS fille FOR SilALL BREAK IN PRESSURIZER 100 m

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200 400 600 800 1000 Time, sec Figure 4 1,473 055 t

69-1106001 SYSTEM PRESSURE VS. TillE.SMALL BREAKS I/O AUXILIARY FEEDWATER 2500 "ZER0" LEAK SMALL 2000 j

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1000 1500 2000 2500 3000 3500 4000 Time, see Figure 5 1473 056.

69-1106001 PRESSURIZER LEVEL VS. TIME-CLASS 3 BREAKS 1/0 AUXILIARY FEEDWATER e

100 p _ __

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0 500 1000 1500 2000 2500 3000 3500 4000 Time, see Figura 8 1473 057

-u-

69-1106001 PABT II - APPENDIX A INADEQUATE CORE COOLING - DESCRIPTION OF PLAhT BEHAVIOR 1.0 INTRODUCTICN Following a loss-of-coolant accident (LOCA) in which the reactor trips, it is necessary to remove the decay heat from the reactor core to prevent da= age. Core heat removal is accomplished by supplying cooling water to the core. The water which is availcble for core cooling is a portion of the initial reactor coolant system (RCS) water inventory plus any water injected by the emergency core cooling system (ECCS). The heat added to the cooling water is removed via the steam generator and/or the break.

As long as the reactor core is hept covered with a mixture of water and steam, core damage will be avoided. If the supply of cooling water to the core is decreased or interrupted, a lower mixture level in the core will result. If the upper portions of the core becomes uncovered, cooling for those regions will be by forced convection to superheated steam which is a low heat transfer regime. Continued operation in the steam cooling mode will result in elevated core temperatures and subsequent core damage.

2.0 LOSS OF RCS Ih"N5 TORY WITH REACTOR COOLANT PNPS OPERATING With the RC pu=ps operating during a small break, the steam and water will remain mixed during the transient. This will result in liquid being discharged out the break continuously. Thus, the fluid in the RCS can evolve to a high void fraction. The void fraction of the RCS 1473 058

(])

({)

69-1106001 indicates the ratio of the volume of steam in the RCS to the total volume of the RCS.

Since the RCS can evolve to a high void fraction for certain small breaks with the RC pumps on, a RC pump trip by any =eans (i.e., loss of offsite power, equipment failure, etc.) at a high void fraction.

during the small break transient may lead to inadequate core cooling.

That is, if the RC pumps trip at a time period when the system void fraction is ;reater than approxi=ately 70%, a core heacup will occur because the amount of water left in the RCS would not be sufficient to keep the core covered. The cladding temperature would increase until core cooling is re-established by the ECC systems.

For certain break sizes and times of RC pump trip, acceptable peak cladding temperatures during the event could not be assured and the core could be damaged. Thus, prompt operator action to trip the RC pu=ps upon receipt of a low pressure ESFAS signal is required in order to ensure that adequate core cooling is provided. Following the RC pu=p rip, the small break transient concerns about inadequate core cooling will be the same as described in the previous section.

If the RC pumps can not be tripped by the operator, the continued forced circulation of fluid throughout the RCS will keep the core cooled.

However, if little or no ECCS is being provided to the RCS, the fluid in the RCS will tentually become pure steam due to the continued enerr,y addition to the fluid provided by the core decay heat. Under these circum-stances, an inadequate core cooling situation will exist.

Since 1473 059

69-1106001 the heat removal process under forced circulation is better than the steam cooling = ode described below for the pumps off situation, the operator actions and indications described in the subsequent section are sufficient for inadequate core cooling with the RC pu=ps operating.

3.0 LOSS OF RCS INVENTORY WITHOUT REACTOR COOLANT PD!PS OPERATING Without the RC pu=ps operating, the cooling of the core is acco=plished As the fluid by keeping the core covered with a steam water mixture.

in the core is heated, some of it or all of it may be turned to steam.

If insufficient cooling water is available to =aintain the steam-water mixture covering the core, the core exit fluid te=peratures will begin to deviate from the saturation te=perature corresponding to the pressure of the RCS. One i==ediate indication that inadequate cote cooling =ay exist in the core is that the te=perature of the core exit ther=occuples and hot leg RTD's are superheated. At this condition inadequate core cooling is evident as the core vill be partially uncovered. However, the degree of uncovery is not severe enough to cause core da= age. This condition is not expected to occur but is not, by itself, a cause for extre=e action. If the ECCS syste=s are functioning nor= ally, the te=peratures should return to satuation without any actions beyond those outlined for a s=all break. For incore ther=occuple te=perature indicating superheated conditions, the operator should (a) verify emergency cooling water is being injected through all HPI no::les into the RCS, (b) initiate any additional sources of cooling water available such as the standby makeup pu=p, (c) verify the stea= generator level is being =aintained the e=ergency level (d) if stea= generator level is not at 957. of at operating range (96 inches indicated on the startup range for raised loop plants), raise level to the 95:: level,

-44 _

69-1106001

, (e) if the desired steam generator level cannot be achieved, actuate any additional available sources of feedwater such as startup auxiliary feedwater pump. (f) establish 100 F/hr. cooldown of RCS via steam generator pressure control, (g) open core flooding line isolation valves if previously isolated, and (h) if RC pressure increases to 2300 psig (1500 psig for DB-1) open the pressurizer PORV to reduce RC pressure and reclose PORV when RC pressure falls to 100 psi above the secondary pressure.

These actions are directed toward depres-surization of the RCS to a pressure at which the ECCS water input exceeds core steam generation.

The alignment of other sources of cooling water is the recognition that the inj ection of the HPI system alone is not sufficient to exceed core boil off.

If the incore thermocouple indications reach curve #1 on Figure 3 in Part I, the peak fuel cladding temperature has reached approximately 14000F.

Above this temperature level there is a potential for cladding, rupture.

Also, the zircaloy cladding water reaction will begin to add a significant amount of heat to the fuel cladding thereby greatly increasing the possibility of core structural damage unless adequate core cooling is restored.

Non-condensible gas formation will increase rapidly from.this level of fuel clad temperature.

For incore thermocouple temperature indications at or exceeding curve

1. 1 oa Figure 3 in Part I, the operator should (a) start one RC pump in each loop, (b) depressurize the steam generator as rapidly as 0

possible to 400 psig or as far as necessary to achieve a 100 F decrease in saturation temperature, (c) immediately continue the plant cooldown by maintaining a 100 F/hr. decrease in secondary saturation temperature to achieve 150 psig RC pressure, (d) open the pressurizer pilot operated relief valve (PORV), as necessary, to relieve RCS pressure and vent non-condensible gases.

The operator action in starting the RC pumps will provide forced flow core cooling and will reduce the fuel cladding temperatures.

The rapid depres-surization of the steam generators will help to depressurice the primary system to the point where the core flooding tanks will actuate.

Stopping the depressuri:ation at 400 psig (or at a 1473 061

69-1106001 reduction in saturation temperature of 100 F) will maintain the tube to shell temperature difference within the 100 F design limit.

The continued cooldown to 150 psig vill reduce the primary system pressure to the point where the Low Eressure Injection System can supply cooling.

The opening of the PORV will also help to depressurize the primary s y s t,e m. - The PORV should be closed when the primary pressure is within 50 psi of the secondary pressure and then should only be used as necessary to maintain the primary system pressure at no greater than 50 psi above the secondary system pressure.

This method of operation will minimize the loss of water from the primary system through the PORV.

If the incore thermocouple readings reach curve #2 on Figure 3 in Part I,

the peak cladding t,emperature is approximately at the 1800 F level.

This is a very serious condition.

At this level of clad temperature, significant amounts of'non-condensible gas are being generated and core damage may be unavoidable.

Extreme measures are required by the operator to prevent major core damage.

The goal of these actions is to depressurize the RCS to a level where the core flooding tanks will fully discharge and the LPI system can be accuated thus providing prompt core recovery.

The operator should (a) depressurize the steam generators as rapidly as possible down to atmospheric pressure, (b) start the remaining RC pumps and (c) open the PORV and leave it open.

4.0 INADEOUATE CORE COOLING RESULTING FROM LOSS OF STFAM GENERATOR HEAT SINK For a very small or non-LOCA event, the core decay heat removal is accomplished via the steam generatcrs.

If that heat removal is de-creased or lost, the natural circulation of fluid within the RCS may be reduced or stopped.

The loss of natural circulation for core 1473 062