ML19257A013

Small Break Operating Guidelines.

ML19257A013
Person / Time
Site:	Davis Besse
Issue date:	11/20/1979
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML19257A008	List: ML19257A007 ML19257A013
References
RTR-NUREG-0578, RTR-NUREG-578 69-1106003, 69-1106003-00, BWNP-20004, BWNP-20004-02, BWNP-20004-2, NUDOCS 7912310486
Download: ML19257A013 (50)

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, i B'M-20004 (6-76)

BABCOCK & WILCOX NUCLEAR POWER GENERATION OlvlSION TECHNICAL DOCUMENT I

DIERGENCY OPERATION SPECIFICATION 69 - 1106003 - 00 Doc. ID - Serial No., Revision No.

for S) FALL BREAK OPERATING GUIDELINES

- DAVIS-BESSE UNIT 1 -

165429h PAGE 1

i BWNP-20006 (6-76)

BABCOCK & WILCOX Nuesta NUCLEAR POWER GENERAtON OlvisiCN 69- ' 6 3-TABLE OF C0i!TEilTS/ EFFECTIVE PAGE l.!ST SEC Tl0ll TITLE PAGE 0 0 0. H 0.

PART I OPERATING GUIDELINES FOR SMALL BREAKS 4 69-1106003-00 1.0 SYMPTOMS AND INDICATIONS (INMEDIATE INDICATIONS) 4 69-1106003-00 2.0 IMMEDIATE ACTIONS 4 69-1106003-00 3.0 PRECAUTIONS S 69-1106003-00 6 69-1106003-00 7 69-1106003-00 8 69-1106003-00 4.0 FOLLONUP ACTIONS 8 69-1106003-00 9 69-1106003-00 10 69-1106003-00 11 69-1106003-00 12 69-1106003-00 13 69-1106003-00 14 69-1106003-00 APPENDIX A LIP COOLING 15 69-1106003-00 16 69-1106003-00 Figure 1 Pressure-Temperature Limit Curve to Preclude Reactor Vessel Brittic Fracture during RCS Depressurization Following Accident Conditions 16-1 69-1106003-00 Figure 2 Minimum Required HPI Flow vs.

RCS Pressure 16-2 69-1106003-00 Figure 3, Core Exit Thermocouple Temperature for Inadequate Core Cooling 16-3 69-1106003-00 PART II SMALL DREAK Pl!ENOMENA - DESCRIPTION OF PLANT BEllAVIOR 17 69-1106003-00

1.0 INTRODUCTION

17 69-1106003-00 2.0 IMPACT OF RC PUMP OPERATION ON A SMALL LOCA 17 69-1106003-00 18 69-1106003-00 3.0 SMALL BREAKS WITil AUXILIARY FEEDWATER 18 69-1106003-00 19 69-1106003-00 20 69-1106003-00 21 69-1106003-00 1654 297 DATE: 11-20 '9 PAGE 3

i BWP-20006 (6-76)

BABCOCK & WILCOX Nuxste NUCLEAA POWER GENERADON CIVIS:ON 69-1106003-00 TABLE OF C0i!TE!!TS/EFFECTi'!E PAGE LIST S ECTIOlt TITLE PAGE 0 0 0. N O.

S.0 SMALL BREAKS WITliOUT AUXILIARY FEEDWATER 21 69-1106003-00 22 69-1106003-00 23 69-1106003-00 6.0 TRANSIENTS THAT >!IGHT INITIATE A LOCA 24 69-1106003-00 7.0 HPI/MU TilROTTLING 24 69-1106003-00 2S 69-1106003-00 FIGURE 1 BREAK SPECTRU.'!-AVERAGE SYSTE31 VOID FRACTION WITil TIIE RC PUMPS OPERATIVE AND 2 IIPI PU31PS 26 69-1106003-00 FIGURE 2 PRESSURE VS TIME-S !ALL BREAKS WITil AUX 1LIARY FEEDWATER 27 69-1106003-00 FIGURE 3 PRESSURIZER LEVEL VS TIME - SMALL BREAKS NITil AUXILIARY FEEDNATER 28 69-1106003-00 FIGURE 4 PRESSURIZER LEVEL VS TISE FOR SSRLL BREAK IN PRESSURIZER 29 69-1105003-00 FIGURE 5 SYSTE5! PRESSURE VS TISS - SSMLL BREAKS W/0 AUXILIARY FEEDNATER 30 69-1105003-0C F1GURE 6 PRESSURIZER LEVEL VS TI E - CLASS 3 BREAKS li/0 AUXILIARY FEEDNATER 31 69-1106003-00 APPENDIX A INADEQUATE CORE C00LI.' G - DESCRIPTION OF PLA';T DEHAVIOR 32 69-1106003-00

2. 0 LOSS OF RCS INVENTORY NITH REACTOR COOLANT PUMPS OPERATING 32 69-1100003-00 33 69-1105003-00 3.O LOSS OF RCS INVENTORY WITHOUT REACTOR COOLANT PUMPS OPERATING 33 69-1106003-00 34 69-1106003-00 4.0 INADEQUATE CORE CCOLING RESULTING FROM LOSS OF STEAM GENERATOR HEAT SINK 3S 69-1106003-00 1654 298 DATE: 11-20-79 PAGE 3-1

3WP-20007 (6-76)

BABCOCK & WILCOX suussa NUCLEAR POWER GENERAficN DIVl31oM TECHNICAL DOCUMENT 6S- 1 6 3- -

PAP.T I - OPERATING GUIDELINES FCR SMALL BREAKS 1.0 SYMPTCMS AND INDICATICNS (I.WEDIATE INDICATIONS) 1.1 Excessive reactor coolant system (RCS) makeup

• 1.2 Decreasing RCS pressure 1.3 Reactor trip 1.4 Decreasing pressuri:er level

• 1.5 ESFAS actuation

• 1.6 Low makeup tank level

• 1.7 Additional criteria during heatup and cooldown*

1.7.1 RCS temperature increasing, minimum letdown and pressuri:er level de creas ing.

1.7.2 With a cooldown of < 100 F/hr and cannot maintain level in makeup tank.

2.0 I.'S'EDIATE ACTIONS 2.1 If the ESFAS has been initiated autcmatically because of low RC pres-sure, ir=ediately secure all RC pumps.

2.2 Verify centrol room indications support the ala =s received, verify autcmatic actions, and carry cut standard post-trip acticas.

2.3 Balance high-pressure injection (HPI) flow between injection lines when HPI is initiated per 3.13.

~ ~

2.4 Monitor system} pressure and temperature. If saturated conditions occur above ESFAS pressure setpoint for initiation of HPI, commence injecting water using both MU pumps into RCS.

2.5 Verify that appropriate Once-Through Steam Generator (OTSG) level is maintained by feedwater control (by ICS control of main feedwater or appropriate level control of auxiliary feedwater).

2.6 If ESFAS has been bypassed due to heatup or cocidown, initiate safety injection.

CAUTION: If 50 F subcooling criteria is met, throttle HPI flow to keep system pressure within nor=al technical spc -i-fication P-T curve limits. If RCS is not 500F subculed, continue full safety injection until 50 0 F subcooling is attained or the P-T limits of Figure 1 are reached.

• May not cecur on all s=all breaks.

1654 299 DATE: 11-20-79 PAGE 4

%WP-20007 (6-76)

BABCOCK & WILCOX NumEn NUCMA3 Powit GENthDoN DM$loN 69-1106003-00 TECHNICAL. DOCUMENT 3.0 PRECAUTIONS 3.1 If the ESFAS has been initiated on low RC presure, termination of RC pump operation takes precedence over all other immediate actions.

NOTE: If ESFAS has been actuated on high RB pressure, then monitor RC pressure and trip RC pu=ps once pressure decreases below the ESFAS low pressure setpoint.

3.2 If ESFAS has been initiated, the RC pump's tripped, and the RCS determined to be at least 50 F subcooled, the operator should establish as quickly as possible if the cause for the depressuri::a-tion is due to either a LOCA or non-LOCA (overcooling) event.

Proceed to step 4.4 for non-LOCA events.

3.3 If the HPI system has actuated because of low pressure conditions, it must remain in operation until one of the following criteria is satisfied:

1. The LPI system is in operation and flowing at a rate in excess of 1000 GPM in each line and the situation has been stable for 20 minutes.

or

2. All hot and cold leg temperatures are at least.50 F below the saturation temperature for the existing RCS pressure and the action is necessary to prevent the indicated pressuri::er level from going off-scale high.

NOTE: If 50 F subcooling cannot be maintained, the HPI shall be reactivated.

NOTE: The degree of subcooling beyond 50 F and the length of time HPI is in operation shall be limited by the pressure /

, temperature considerations for the vessel integrity (see Section 3.4).

3.4 When the reactor coolant is > 50 F subcooled, the reactor vessel downcomer pressure / temperature (P-T) combination shall be kept below ,

and to the right of the limit curve shown in Figure 1. The downecmer temperature shall bt: determined as follows:

3.4.1 With one or more RC pumps operating use any cold leg RTD as an indica-tion of reactor vessel downcomer temperature.

1654 300 DATE: 11-20-79 PAGE 5

BWP-20007 (6-76)

BABCOCK & WILCOX Numsta NUC11Aa PowEt otHERAT1oM DIVt51oM 69- 1 6003-00 TECHNICAL. DOCUMENT 3.4.2 With no RC pumps operating the RV downcomer temperature shall be determined by averaging the five lowest incere thermocouple temperature readings and subtracting 150 F from the average incore thermocouple temperature value.

5 IT T - 150 F DWN

• 5 where TDWN = average RV downcomer temperature, F 5

IT e

= sum of the 5 lowest incore thermocouple temperature readings.

NOTE: Figure 1 is applicable only under LOCA conditions, The P/T curve in the technical specification is valid for all other operating conditions.

NOTE: When the reactor coolant is less than 50 F subcooled, the reactor vessel downcomer pressure temperature combination will inherently be below and to the right of the limit curve.

Therefore, no operator action will be required to prevent exceeding the reactor vessel integrity limits until after a > 50 0F subcooled margin exists.

NOTE: When the reactor coolant is > 50 F subcooled, RC pressure can be reduced by reducing tee HPI flow rate to avoid exceeding the RV integrity limits.

3.5 Pressurizer level may be increasing due to RCS reaching saturated conditions or a break on top of the pressurizer.

3.6 If high activity is detected in a steam generator, isolate the leaking generator. It is recocmended that both steam generators not be isolated.

3.7 Other indications wh!.h can confirm the existence of a LOCA:

3.7.1 RC drain tank (quench tank) pressure (rupture disk may be blown) .

3.7.2 Increasing reactor building sump level.

3.7.3 Increasing reactor building temperature.

3.7.4 Increasing reactor building pressure.

3.7.5 Increasing radiation monitor readings inside containment.

3.7.6 Reactor coolant system temperature becoming saturated relative to the RCS pressure.

DATE:

1654 301 PAG-c 6

Bk'NP-20007 (6-76)

BABCOCK & WILCOX Nu,4:st NUCLEAR Powlt otNtRAfloN Olvi$loN 69-1106003-00 TECHNICAL. DOCUMENT 3.7.7 Hot leg temperature equals or exceeds pressuri:er temperature.

3.7.8 Increase in the excore neutron ' detector indications.

NOTE: In conjunction with the indications in 3.10.1, this could be an indication of inadequate core cooling.

3.8 HPI cooling requirements could deplete the boratqd water storage tank, and initiation of LPI flow from the reactor building sump to the HPI pumps would be required.

3.9 Alternate instrument channels should be checked as available to confirm key parameter readings (i.e., system temperatures, pressures and pressuri:er level) .

3.10 Maintain a temperature versus time plot and a corresponding temperature pressure plot on a saturation diagram. Using hot leg RTD's and highest incore thermocouple reading, these plots will make it possible to track the plant's condition through plant cooldown. -

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 thermocouple temperature reads superheated for the existing RCS pressure.

3.10.2 If primary temperature and pressure indications correspond to saturated coolant conditions prior to or during plant cooldown, a return to subcooled coolant conditions may occur.

3.11 Component cooling water (CCW) and seal injection should be maintained to the RC pumps to insure continued service or the ability to restart the pumps at a later time.

3.11.1 Normal limits and precautions apply for RC pump operation.

3.11.2 If the RC pumps are tripped for any reason, seal injection should be maintained to ensure long term seal integrity.

3.12 If RCS pressure increases above the shutoff head of the HPI pumps, the MU system should be utilized and controlled per Section 3.3.

3.13 If action is necessary to balance HPI flow, then the following steps should be taken:

1. If both HPI trains are available, balance flow between lines in each train.

2. If only one HPI train is available, throttle the high flow line down to but not below Figure 2.

DATE: 2- PAGE 11-20-79 7

, . SWNP-20007 (6-76)

BABCOCK & WILCOX NuusEn NUCLEAR power GtNERAfloN DIVI $loN 69-1106003 TECHNICAL 000Uf.1ENT -

3.14 If action is necessary to balance LPI flow, then the following steps should be taken:

1. If both LPI trains are available, throttle the control valves as necessary to balance LPI flow.

If only one LPI train is available, open the LPI cross-connect 2.

and balance flows. Throttle the control valves as necessary to prevent pump runout, per plant limits and precautions.

4.0 FOLLOWUP ACTIONS 4.1 Identification and Early Control

'4 .1.1 If HPI has initiated because of low pressure, control HPI in accordance with step 3.3.

4.1.2 If both HPI trains have not actuated on ESFAS signal, start second HPI train if possible. Balance HPI flows.

4.1.3 If RC pressure decreases continuously, verify that core flood tanks (CFTs) and low pressure injection (LPI) have actuated as needed, and balance LPI.

4.1.4 If cause for cooldown/depressurization is determined to be due to a non-LOCA overcooling event and the RCS is at least 50 F subcooled then proceed'to section 4.4.

4.1.5 Attempt to locate and isolate leak if possible. Letdown was isolated in step 2.2. Other isolatable leaks are PORV (close block valve) and between valves in spray line (close spray and block valve).

4.1.6 Determine availability of reactor coolant pumps (RCPs) and main and auxiliary feedwater systems. If feedwater is not available, go to 4.2. If feedwater is available, go to 4.3.

4.2 Actions if Feedwater is not Available 4.2.1 Throughout the following steps maintain maximum HPI flow and restore feedwater as soon as possible. The electric startup feed-water pump is an alternate source of feedwater, and shculd be aligned and started as soon as possible.

4.2.1.1 If ESFAS has actuated due to low pressure, maintain steam generator level at 96 inches on the SU range instrumentation. If low pressure ESFAS has not actuated maintain level at greater than 35 inches.

4.2.2 If RCPs are operating, go to one pump per loop. If RCPs are not '

operating, go to step 4.2.5 below.

4.2.3 If RCS pressure increases above the HPI shutoff head, open FORV and leave open: align / actuate 50 system per 3.12 and control per HPI instructions given in 3.3.

DATE: ^

11-20-79 g

BWP-20007 (6-76)

BABCOCK & WILCOX Numan NucteAn rown GeNHATION Om$10N 69-1106003-00 TECHNICAL DOCUMENT NOTE: If the PORV cannot be actuated, the safeties will relieve pressure.

4.2.4 When auxiliary feedwater is recovered, restore OTSG 1evels in a controlled manner and establish appropriate OTSG 1evel control.

Secure electric Su-FW pump if operating and close PORV or block valve, if possible. Proceed to step 4.3. 2.

4.2.5 If no RCPs are operating, open PORV, maintain HPI flow, and align / actuate MJ system per 3.12.

NOTE: If the PORV cannot be actuated, the safeties will relieve pressure.

4.2.6 When auxiliary feedwater flow is restored, resume automatic OTSG level control at 96-inches indicated on the SU-range instrumentation and close PORV or block valve, if possible.

4.2.7 Verify natural circulation in the RCS by observing:

4.2.7.1 Cold leg temperature is saturation temperature of secondary side pressure within approximately 5 minutes.

4.2.7.2 Primary AT (EOT-TCOLD) becomes constant.

4.2.8 .Go to step 4.3.4.1.

4.3 Actions with Feedwater Available to One or Both Generators 4.3.1 Maintain ene RCP running per icop (stop other RCPs) . 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 cooling by adjusting steam pressure via turbine bypass and/or atmospheric dumps. Cooldown at 100 F per hour to achieve an RC pressure of 250 psig. Refer to pre-caution 3.10 for development of temperature and pressure plots.

Isolate core flood tanks when 50 F subcooling is attained and RC pressure is less than 700 PSIG. Go into LPI cooling per Appendix A.

4.3.4 If RCPs are not operating:

4.3.4.1 Verify steam geneutor level control at appropriate setpoint and verify the conditions in step 4.2.7.

)

1654 304 DATE: 11-20-79 PAGE 9

BWP-20007 (6-76)

BABCOCK & WILCOX NuusEn NUCLEAR POWER GENERATION OtVl$loN 69-1106003-00 TECHNICAL DOCUMENT 4.3.4.2 If RC pressure is decreasing, wait until it stabili::es or begins increasing. If it begins increasing, go to step 4.3.4.4.

4.3.4.3 Proceed with a controlled cooldown at 100 F/hr by controlling steam generator secondary side pressure. Monitor RC pressures and tempera-tures during cooldown and proceed as indicated below:

4.3.4.3.1 If RC pressure continues to decrease, following secondary side pressure decreases and with primary system temperatures indicating saturated conditions, continue cooldown until an RC pressure of 150 psi is reached, and proceed to step A.4 of Appendix A.

4.3.4.3.2 If RC pressure stops decreasing in response to secondary side pressure decrease and reactor system becomes subcooled, check to see that the following conditions are both satisfied:

A) All hot and cold leg temperatures are below the saturation temperature for the existing RCf pressure, and B) The hot and cold leg temperatures are decreasing in response to steam generator secondary temperature decrease.

If these conditions are satisfied and remain satisfied, gontinue cooldown to achieve an RCS temperature (cold leg) of 280 F, and pro-ceed to step A.1 of Appendix A.

NOTE: If the conditions above are met below 700 PSIG, the core flood tanks should be isolated.

NOTE: If the primary system is 50 F subcooled in both hot and cold legs and primary system pressure is above 250 PSIG, starting a reactor coolant pung is permissible. If system does not return to at least 50 F subcooling in two minutes, trip pumps.

If forced circulation is achieved, proceed to step 4.3.

4.3.4.3.3 If RC pressure stops 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 possible per the instructions below. If RC pumps 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 than 1600 PSIG.

1654 305 DATE: 11-20-79 PAGE 10

BWP-20007 (6-76)

BABCOCK & WILCOX Numsen NucteAn Powen ceNenAnos civision 69-1106003-00 TECHNICAL. DOCUMENT 4.3.4.4.2 If reactor coolant system pressure exceeds steam generator secondary pressure by 600 PSIG or more " bump" one reactor coolant pump for a period of approximately 10 seconds (preferably in operable steam generator loop). Allow reactor coolant system pressure to stabili:e.

Continue cooldown. If reactor coolant system pressure again exceeds secondary pressure by 600 psi, wait at least 15 minutes and repeat the pump " bump". Bump alternate pumps so that no pump is bumped more than once in an hour. This may be repeated, with an interval of 15 minutes, up to 5 times. After the fifth " bump", allow the reactor coolant pump to continue in operation.

4.3.4.4.3 If pressure has stabili:ed for greater than one hour, secondary pressure is less than 100 PSIG and primary pressure is greater than 2SO PSIG, bump a pump, wait 30 minutes, and start an alternate pump.

4.3.4.5 If forced flow is established, go to step 4.3.3.

4.3.4.6 If a reactor coolant pump cannot be operated and reactor coolant system 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 PORV is not operable, pressuri:er safety valves will relieve overpressure.

4.s.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 generator cooling is established as indicated by establishing con-ditions 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.

4.4 Non-LOCA Overcooling Transient with Feedwater Available, 4.4.1 Immediately restart a RC pump in each loop if the RCS is 50 F subcooled.

4.4.2 Control steam pressure via turbine bypass or atmospheric dump valves to stabili:e or control plant heatup.

N ME: Considerable HPI may have been added to the RCS. Therefore, to prevent RCS from going solid, the above action may be necessary.

4.4.3 As long as the RCS is maintained 50 F subcooled, throttle HPI/MU and letdown flow to maintain pressuri:er level at N 100 inches. ,

4.4.4 Using turbine bypass valves and feedwater system, control steam generators as needed to limit plant hea+up until RC pressure control can be re-established with the pressuri er.

NME: Cold RCS water may have been added to the pressurizer; there-fore, a period of time may elapse before normal RC pressure control can be established with the pressuri:er heaters.

1654 306 DATE: 11-20-79 PAGE 11

BWNP-20007 (6-76)

BABCOCK & WILCOX Nuuse NUCLEAR PCwth GENERATION DIVISION 69-1106003-00 TECHNICAL. DOCUMENT 4.4.5 Once pressure control is re-established, use nomal heatup/cooldown procedure to establish desired plant conditions.

4.5 Actions for Inadecuate Core Cooling 4.5.1 I= mediate steps for inadequate core cooling NOTE: If RC pumps are mnning, do not trip pumps. This supercedes instmetions in section 2.1.

4.5.1.1 Verify HPI syste=s are functioning properly with maximum flow. St art makeup pu=p(s), if possible, to increase injecticn 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.

CAUTICN: Reference leg boiling could give false level indication.

4.5.1.3 Depressuri:e operative steam generator (s) to establish a 100 F/hr decrease in secondary saturation temperature.

4.5.1.4 Ensure core flood tank isolation valves are open and that LPI actuates if pressure reaches 420 psi.

4.5.1.5 If reactor coolant system pressure increases to 2300 PSIG (1500 PSIG for DB-1) open pressurizer PORV to reduce reactor coolant system pressure. Reclose PORV when RCS falls to 100 PSIG above the secondary pressure. Repeat if necessary. If PORV is not operable, pressuri:er safety valves will relieve pressure.

4.5.1.6 Proceed i= mediately to 4.5.2.

4.5.2 When the indicated incore ther=occuple temperatures or hot leg RTD te=peratures are superheated for the existing RCS pressure, operator action shall be based en conditions detemined frem Figure 3, by a sa=ple of the highest incere themoccuple temperature readings to detemine the core exit themoccuple tenperature.

NOTE: More than one thex=occuple temperature reading should be used (for example use the average of 5) .

4.5.3 When the incere themoccuple te=perature has been determined per Section 4.5.2, go to the section indicated below.

DATE: 11-20-79 AGE y2

BWP-20007 (6-76)

BABCOCK & WILCOX Numsen NucteAn powen ceNeRADON Om33oN 69-1106003-00 TECHNICAL 00C0 MENT Incore Thermocouple Tercerature Section Incore Tc 1 Saturation 4.1.6 Curve 1 i Incore Tc < Curve 2 Figure 3 4.5.4

~

Incore Tc > Curve 2 Figure 3

, 4.5.5 The incore thermocouple temperature readings shall be continuously

' NOTE *'

monitored. If the t-emperature is between. saturation and Curve-1 Figure 3, only the preceeding actions will be taken until the indicated incore thermocouple temperatures return to saturation temperature for the existing RCS pressure or the temperature increase t Curve 1 Fi ure 3.

,4.5.4 Ac'. ions for curve ,1 1 incore Tc < Curve 2 Figure 3 4.5.4.1 If RC pumps are not operating, start one pump per loop (if possible) .

This instruction supersedes previous instructions to trip RC purps.

NOTE: Do not bypass normal interlocks.

4.5.4.2 Depressuri:eroperativesteamgenerator(s)asrapidfyaspossible to 400 PSIG or as far as necessary to achieve a 100 F decrease in secondary saturation temperature.

4.5.4.3 Open the PORV, as necessary, to maintain RCS pressure within 50 psi of steam generator secondary side pressure.

NOTE: If steam generator depressuri:ation was not possible, open PORV and leave open.

4.5.4.4 Immediately continue plant cooldown by maintaining 100 F/hr.

Decrease in secondary saturation temperature to achieve 150 PSIG RCS pressure.

CAUTION: If auxiliary feed pu:rp is supplied by main steam, do not decrease pressure below 50 psia for auxiliary feed pu=p operation.

4.5.4.5 If the average incore ther=ocouple temperature increases to Curve 2 Figure 3 proceed immediately to Section 4.5.5.

4.5.4.6 When RCS pressure reaches 150 PSIG, go to Appendix "A".

. 4.5.5 Actions for Incore Tc > Curve 2 Figure 3 4.5.5.1 If possible, start all RC pumps.

NOTE: Starting interlocks should be defeated if necessary, but in order to minimi:e the possibility of a fire due to bypassing seme interlocks, the following precautions should be observed:

1) Do not defeat the overload trip circuit and 2) if CCW is not restored to the motor within 30 minutes, trip the RC pump.

DATE: 11-20-79 ^

13

SWNP-20007 (6-76)

BABCOCK & NVILCOX NUM8it NUCttAS PoWit QtN(RA ';CN OlVi$10N 68-22 6 3-TECH11 CAL DOCUf1EllT It sheuld be recegni:ed that starting the RC pumps withcut cooling and/e' injection water will probably fail the pt:mp sesis and m. cause the pu=p shaft to break. How-ever, some core cooling will be provided prior to destructicn of the pu=p. Breakage of the pump shaft will not cause

consequential damage cutside of the pump.

4.5.5.2 Depressuri:e the,cperative stess generator (s) as quickly as possible to between 50 to 75 psia.

CAUTION: If auxiliary feed pump is supplied by main steam, do not decrease pressure below 50 psia or steam driven auxiliary feedwater pu=p will not deliver water to the steam generators.

4.5.5.3 Open the pressuri:er PORV and leave open.

NOTE: The RCS will depressuri:e and the LPI system should restore core cotiing.

4.5.5.4 When incore dier=occuple temperatures return to the saturation temperature for the existing RCS pressure; and the LPI system is delivering flow, proceed as follows:

4.5.5.4.1 Close the pressuri:er PORV; reopen if RCS pressure increases above 150 PSIG.

4.5.5.4/2 Decre tse to two (2) RC pump operation (one per loop).

4.5.5.4.3 Isolate the core flood tanks.

4.5.5.4.4 Faintain steam generator pressure between 50 and 75 psia. Do not lower below 50 psia for auxiliary feed pump operation.

4.5.5.4.5 Control HPI per 3.3.

4.5.5.4.6 When BWST level reaches LO-LO level limit, verify Tutomatic function s have switched LPI suction to RB sump and LPI suctran from BWST is c]osed.

NOTE: If HPI is required per 3.3, align LPI and HPI in piggyback mode. Shut off minimum recirculation lines to BWST.

CAUTION: Do not deadhead HPI pumps when minimum recirculation line s are closed. This may damage pumps.

4.5.5.4.7 Go to Appendix "A".

1654 309 Date: 11-20-79 Page 14

3WP-20007 (6-76)

BABCOCK & WILCOX Nuuset NUCLEAR PCWER GENERAtloN Olvt$loN TECHNICAL DOCUMENT 69-1106003-00 APPENDIX A LPI COOLING A.1 Determine if primary coolant is at least 50 F subcooled. If not, go to step A.3.

A.1.1 Start LPI pumps. If both pumps are operable, go to step A.2. For one LPI pump operable maintain CTSG cooling and proceed as follows.

The operable LPI pump will be used to maintain system inventory.

A.1.2 Obtain primary system conditions of < 280 F and < 250 PSIG.

A.1.3 Align the discharge of the operable LPI pump to the suctions of the HPI pumps and take suction from the BWST. If the BWST is at the low level alarm, align LPI suction from the RB svl::p and shut suction from BWST.

A.1.4 Start the operable LPI pump specified above. The HPI-LPI systems will now be in " piggy back" and HPI flow is maintaining system pressure.

A.1.5 Go to single RC pump operation.

A.1.6 When the second LPI pump is available, align it in the decay heat mode and commence decay heat removal. (Decay heat system flow greater than 1000 GPM). Secure remaining RC pump when decay heat removal is established.

CAlTTION: Verify that adequate NPSH exists for the decay heat pump in the DH removal mode. If inadequate, transfer to LPI mode.

A.1.7 Reduce reactor coolant pressure to 150 PSIG by throttling HPI flow.

Control RC temperature using the decay heat system cooler bypass to maintain system pressure at least 50 psi above saturation pressure, to assure that NPSH requirements for the decay heat pump are maintained.

A.1.8 Secure the HPI pump and shift the LPI pump supplying it to the LPI injection mode.

A.1.9 Reduce reactor coolant temperature to 100 F by controlling the decay heat system cooler bypass.

NOTE: If one of the LPI/ decay heat pumps is lost, return to OTSG cooling using natural circulation or one reactor coolant pump.

Go to A.1.1.

A.2 goldown on Two LPI Pumus A.2.1 Maintain RCS pressure at 1 250 PSIG and reduce RCS temperature to 1 280 F.

A.2.2 Align one LPI pump in the decay heat removal mode.

DATE: 11-20-79 "^ " 5 0

3WNP-20007 (6-76)

BABCOCK & WILCOX Nuus tR NUCLEAA POWER otNERAfloN QiVl$3oN 69-1106003-00 TECHillCAl. 000UMEllT A.2.3 Secure one RC pump if two are operating.

A.2.4 Start the decay heat pump in the decay heat removal mode, and when decay heat system flow is greater than 1000 GPM, secure the running RC pump.

A.2.5 Reduce RC pressure to 150 PSIG by throttling HPI flow. Control RC temperature to maintain at least 50 psi margin to saturation pressure.

A.2.6 Start the second LFI pump in the LPI injection mode. Secure HPI pump.

A.2.7 When BWST level reac':;es LO-LO level limit, verify automatic functions have swi~tche~d LPI suction to RB sump and LPI suction from BWST is closed.

A.2.8 Reduce reactor coolant temperature to 100 F by controlling the decay heat system cooler bypass.

NOTE: If one of the LPI/ decay heat pumps is lost, return to OTSG cooling using natural circulation or one RC pump. Go to A.1.1.

A.3 Cool Down RC System at Saturation A.3.1 Maintain RC pressure at < 250 PSIG.

A.3.2 Align one LPI pump to suction of the HPI pumps and the suction to the reactor building sump. (Shut BNST suction valve for this pump.)

A.3.3 When the BWST level reaches the lo-lo level limits, start the LPI pump and shut the HPI pu=p suction from the BWST.

A.3.4 When primary system temperature becomes subcooled by at least 50 F, go to A.1.1.

A.4 Cooldown without Reactor Coolant Pumos A.4.1 RCS initial conditions are: pressure 150 psi, temperature at satura-tion.

A.4.2 Align low pressure injection system for suction from reacter building sump and place into service.

A.4.3 Balance LPI injection and control RC temperature with decay heat coolers.

A.4.4 Isolate core flood tanks.

A.4.5 Go to step A.1.1 and follow the procedure given there, ignoring the instruccions relating to RC pump operation.

DATE: 11-20-79

) f)}f )] PAGE 16

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B'4NP-20007 (6-76)

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P00R ORGINAL 1654 313 DATE: 11-20-79 PAGE 16-2

DUNP-20007 (6-76)

BABCOCK & WILCOX 14UCtf AR PCWit GENERAllCN DtVi$ TON 69- 6 5-TECl!!!!C/tl DOCU . lei!T Figure 3 Core Exit Thermocouple Temperature for Inadequate Core Cooling 1200 -

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?00RBRGINM 1654 314 DATE: 11-20-79 PAGE 16-3

BWNP-20007 (6-76)

BASCOCK & WILCOX Numan sucteAn powen ceNehDoN OmsloN 69-1106003-00 TECHNICAL. DOCUMENT PART II: SMALL BREAK.PHENCMENA - DESCRIPTION OF PLANT BEHAVIOR

1.0 INTRODUCTION

A loss-of-coolant accident is a condition '.i which liquid inventory is lost from the reactor coolant system. Due to the loss of mass from th.e reactor coolant system, the most significant short-term sy=ptom of a loss-of-coolant accident is an uncontrolled reduction in the reactor coolant system pressure. The reactor protection system is designed to trip the reactor en low pressure. This should occur before the reactor coolant system reaches satu-ation 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 systems are designed to respond automatically to low reactor coolant pressure conditions and take the initial actions to protect the reactor core. They are si:ed 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 pre-sented 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.

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 pressuri:er 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.

2.0 IMPACT OF RC PUMP OPERATION ON A SMALL LOCA 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 si:e and location. Continued RC pump operation, even at high system void fractions, will provide sufficient core flow to keep cladding temperatures within a few degrees of the saturated fluid temperature.

DATE:

11-20-79 1654 315 gA0E l e

BWP-20007 (6-76)

BABCOCK & WILCOX Nuw R NUCLEAR POWER otNERATioN Divl51oM 68- 6003-00 TECHNICAL DOCUMENT Since the RCS can evolve to a high void fraction for certain small breaks with the RC pups 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 pu=ps 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, ne cladding temperature would increase until core cooling is re-established by the ECC systems. For certain break si:es and times of RC pump trip, acceptable peak cladding temper-atures during the event could not be assured and the core could be damaged. Thus, prompt operator action to trip the RC pumps upon receipt of a low pressure ESFAS signal is required in order 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.

3.0 ST!ALL BREAKS WITH AUXILIARY FEEDWATER There are four basic classes of break response for small breaks with auxiliary feedwater. These are:

1. LOCA large enough to depressuri:e the reactor coolant system

2. LOCA which stabili:es at approximately secondary side pressure

3. LOCA which may repressuri:e in a saturated condition 4 Small LOCA which stabili:es 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 Deuressuri:e Reactor Coolant System Curves 1 and 2 of Figure 2 show the response of RCS pressure to breaks that are large enough in combination with the ECCS to de-pressuri:e 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 -how the pressuri:er level transient. Rapidly falling pressure en.ses the hot legs to saturate quickly. Cold leg temperature reaches saturation somewhat later as RC pumps coast down or the RCS depressuri:es 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. Upon receipt of a low pressure ESFAS signal, the operator must trip all RC pumps and vetrify that all ESFAS actions have been completed.

The operator mast. also verify that the HPI train flow is balanced through each inj wtion line.

1654 316 DATE. 11 23 79 PAGE

BWP-20007 (6-76)

BABCOCK & WILCOX Nuusen wucttAa powen GENERAfloN OlVl$1oM TECHNICAL DOCUMENT The operator should also balance LPI flows, should 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 condition. Rapid depressuri:ation of the steam generators would only act to accelerate RCS depressuri:ation. 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 pump suction to the reactor building sump.

3.2 LOCA which Stabili:es at Aporoximately Secondary Side Pressure Curve 3 of Figure 2 shows the pressure transient for a break which is too small in combination with the operating HPI to depressuri:e the RCS. The steam generators are, therefore, necessary to remove a portion of core decay heat. Although the system pressure will initially stabili:e near the secondary side pressure, RCS pressure may eventually begin falling as the decay heat level decreases. Curve 3 of Figure 3 shows pressuri:er level behavior. The hot leg temperature quickly equali:es to the saturated temperature of the secondary side and controls primary system pressure at saturation. The cold leg tempera-ture may remain slightly subcooled. If the HPI refills and repressuritu the RCS, the hot legs can become subcooled. The immediate operator action is to trip the RC paps upon receipt of the low pressure ESFAS signal and then verify ESFAS functions. The operator must then verify that the HPI train flow is alanced through each injection line.

Followup action by the operator is to verify OTSG 1evel control at 96 inches on SU range and check for established natural circulation.

This is done by gradually depressurizing the steam generators. If this test fails, intermittent bumping of a RC pump should be performed as soon as one is available. Continued depressuri:ation of .

steam generators with natural circulation leads to cooling and depressuri:ation of the RCS. The operator's goal is to depressuri:e 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.

3.3 LOCA which may Repressuri:e in a Saturated Condition Curve 4 of Figure 2 shows the behavior of a small break that is too small, in combination with the HPI, to depressurize the primary system.

Although steam generator feedwater is available, the loss of prim ry system coolant and the resultant RCS voijing will eventually lead to interruption of natural circulation. This is followed by gradual repressurization of the pri=ary system. It is possible that the primary system could repressuri:e as high as the pressuri:er 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 direct steam condensation in the regions of the steam generators contacting secondary side coolant, the primary system is forced to depressuri e to the saturation pressure of the secondary side.

11-20-79 1654 317 19

BWNP-20007 (6-76)

BABCOCK & WILCOX NU"III NUCLEAA Powit GENERATION OtVtSioN 69- o6003-o0 TECHNICAL DOCUMENT Since the cooling capabilities of the secondary side are needed to continue to remove decay heat, RCS pressure will not fall below that on the secondary side. HPI flow is sufficient to replace the inventory lost to boiling in the core, and condensation in the steam generators removes decay heat energy. The RCS is in a stable thermal condition and it will remain there until the cperator takes further action. The pressurizer level response is characteri:ed by Curve 3 of Figure 3 during the depressuri:ation, and Curve 4 of Figure 3 during the temporary repressuri:ation phase. The dashed line indicates the level behavior if pressure is forced up to the pressuri:er safety valve setpoint. During this transient, hot leg temperature will rapidly approach saturation with the initial systen depressuri:ation, and it will remain saturated during the whole transient. Cold leg temperature will approach saturation as circula-tion is lost, but may remain slightly subcooled during the repressuriza-tion phase of the transient. Later RCS depressuri:ation could cause the cold leg temperatures to reach saturation. Subsequent refilling of the primary system by the HPI might cause temporary interruption of steam condensation in the steam generator 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 will 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:ence the HPI and break can balance decay heat or the operator takes some action.

The operator's immediate 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 verify steam generator level control at 96 inches indicated and check for natural circulation. If it is positive, he should depressuri:e the steam generators, cool and depressuri:e 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 depressuri:ing the RCS with the steam generators and establish long-term cooling.

3.4 Small LOCA which Stabili:es at P > Psee Curve S 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 pressuri:er has emptied. The primary system remains subcooled and natural circulation to the steam generator removes core decay heat. The pressuri:er never e=pties and continues to control primary system pressure. The operator needs to trip the RC pumps and ensure that ESFAS actions have occurred. Followup actions include pressuri:er inventory control after the RCS is 50 F subcooled and verification that natural circulation is established if the RC pumps are tripped. A restart of the RC pumps under these conditions is desirable for plant control.

DATE: 11-20-79 1654 318 e^cE 20

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BABCOCK & WILCOX su m en NUCttAR Powte GENERAUoN OtVI$ ion 69-1106003-00 TECHNICAL DOCUMENT 3.5 Small Breaks in Pressuri:er The system pressure transient for a small break in the pressuri:er will behave in a manner similar to that previously discussed. The initial depressuri:ation, however, will be more rapid as the initial inventory loss is entirely steam.

The pressuri:er level response for these accidents will initially behave like a very small break without auxiliary feedwater. The initial rise in pressuri:er level shown in ?igure 4 will occur due to the pressure reduction in the pressuri:er and an insurge of coolant into the pressurizer from the RCS. Once the reactor trips, system contraction causes a decreasing level in the pressuri:er. Flashing will ultimately occur in the hot leg piping and cause an insurge into the pressuri:er. This ultimately fills the pressurizer. For the remainder of the transient, the pressuri:er will remain full.

Toward the later stages of the transient, the pressuri:er may contain a two-phase mixture and the indicated level will show that ,

the pressuri:er is only partially full. Except for closing the PORV block val e, operator actions and system response are the same for these brea!-; as for similar breaks in the loops.

4.0 SMALL BREAKS WI'iMOUT 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 repressuri:ation.

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

4.1 LOCA's Large Enough to Depressuri:e Reactor Coolant System 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-term stabilization of these breaks. The pressuri:er 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.

1654 319 DATE: 11-20-79 21

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BABCOCK & WILCOX NU"III NUCLEAR Powl4 oENERAfloN OlVl5&oN 69-1106003-00 TECHNICAL DOCUMENT 4.2 LOCA's Which Reach a Semi-Stabilized State For Class 2 (Curve 2 of Figure S) breaks, the RC pressure will rapidly reach the low pressure ESFAS trip signal (about twc to three minutes) .

With the HPI's on, a slow system depressuri:ation will be established coincident with the decrease in core decay heat. No CFT actuation is expected. Auxiliary feedwater is not necessary for the short-term stabilization of these breaks. The pressuri er level for this tran-sient rapidly falls off scale.

The operator needs to trip the RC pumps upon the low pressure ESFAS signal, verify completion of all ESFAS functions, and try to establish secondary side cooling. Balancing of the HPI must also be performed.

If steam generator feedwater cannot be obtained and RCS pressure is increasing, the operator should open the PORV and provide all the HPI and makeup capab,ility possible. The goal is to depressuri:e and coc1 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-term cooling established.

4.3 Small LOCA's Which do not Actuate the ESFAS Automatic ESFAS actuation will not occur for Class 3 (Curve 3 of Figure S) 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 repressuri:ation to the PORV or the pressuri:er safety valves will occur for smaller breaks in this class. For the ":ero" break case, repressurization to the PORV will occur in the first five minutes. Operator action is re-quired within the first 20 minutes to ensure core coverage throughout the .ransient. For DB-1, this action is manual actuation of auxiliary feedwater for a small break. For a ":ero" break, the combined use of the 3RJ system, startup feedwater pump, and PORV at high RC pressures (above the shutfoff head of the HPI) has also been shown to provide acceptable results. In all cases, AFW should be restored as quickly as possible.

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, pressuri:er level response will be as shown in Figure 6. The minimum refill time for the pressuri:er is that for the ":ero" break and is shown in Figure 6. After initially drawing inventory from the pressuri er, the system repressurization will cause the pressuri:er level to increase, possibly to full pressuri:er level.

Once the operator action to restore auxiliary feedwater has been taken, the system depressurization will result and cause an outsurge from the pressurizer. Complete loss of pressuri:er level may result. For the smaller breaks in Class 3 which result in a system repressurization following the actuation of the HPI system, pressurizer level will increase and then stabili:e.

DATE: 654 320 e^ce 11-20-79 22

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BABCOCK & WILCOX ~umen nuctaAe ROWER otNERADoN DM$loN 69-1106003-00 TECHNICAL DOCUMENT Without auxiliary feedwater, Sth the hot and cold leg temperatures will saturate early in the transient and, for the Class 1 and 2 breaks, will remain saturated. For the Class 3 breaks, once auxiliary feed-water is established, the cold leg temperatures will rapidly decrease to approximately the saturation temperawre corresponding to the SG secondary side pressure and will remain there throughout the remainder of the transient. Hot leg temperatures will remain saturated throughout the event. The operator needs to verify automatic actuation of all ESFAS actions on low pressure, balance HPI flow and attempt to restore secondary side cooling. In the meantime, he should actuate the startup feedwater pump and makeup pumps and open the PORY (if pressure increases) in order to cool the core and limit the RCS repressurization. Once feedwater is available, he can close the PORV and continue the RCS cooldown and depressuri:ation with the steam generators. If natural circulation has not been established, he can bump a RC pump to cause forced circulation. The goal is to depressuri:e 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 pressurizer with feedwater.

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

S.0 TRANSIENTS WITH INITIAL RESPONSE SIMILAR TO A SMALL BREAK Several 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 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 alarms for the affected steam generator, and steam generator overfills result in high steam generator level indica-tions. The overcooling transients will repressurize the primary system and will result in a subcooled condition during repressurization. The immediate actions for both overcooling and small break transients are the same, including tripping of the RC pumps.

The operator will recogni:e overcooling events during repressuri:ation, if not sooner, and is instructed to 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 alarm but does not give an ESFAS actuation alarm.

DATE: 11-20-79 PAGE 23 I

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BABCOCK & WILCOX uunta suct:Ae Powet otNtAAnoN OfVl$loN 69-1106003-00 TECHNICAL DOCUMENT 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. Steam generator levels remain high and the system becomes subcooled during repressuri:ation.

Design features of the B5W NSS provide automatic protection during the early part of sna11 break transients, thereby providing sdequate time for small breaks to be identified and appropriate actica 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.

6.0 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 will not actuate during anticipated transients such as loss of main feedwater, turbine trip, or loss of offsite power.

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

o ESFAS actuation o Quench tank pressure / temperature alarms o Saturated primary system -

o Rising pressuri:er level o High quench tank level o High containment sump level o High containment pressure 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 PORV occur after a transient, they can be identified by initially decreasing RCS pressure and convergence to saturation conditions in the reactor coolant. Small break repressurization, if it occurs, will follow saturation conditions.

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

7.0 HPI/MU THROTTLING For small LOCA's, the HPI/>U system is needta to provide makeup to the RCS and must remain operable unless specific criteria are satisfied.

The basis for these criteria are described below.

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BA8 COCK & WILCOX sumu NUCLEAR POWER oENERAfloN DIVl$loN 69-1106003-00 TECHNICAL. DOCUMENT For certain small breaks, system depressuri:ation will result in LPI actuation. Since the LPI is designed to provide injection at a greater capacity than the HPI, termination of the HPI is allcwed.

However, this action should only be taken if the flow rate through each line is at least 1000 gym and the situation has been stable for 20 minutes. The 20-minute time delay is included to ensure that the system will not repressuri:e 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 sufficient to remove decay heat and ensures that upon temination of the HPI pumps, adequate flow is being delivered to the RV.

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

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 primary system is solid. Continued HPIStU flow at full capacity may result in a solid pressurizer and continued makeup would result in a lifting of the PORV and/or the pressurizer code safety valves. This-cay in turn lead to a LOCA. Thus, flow should be throttled to maintain a stable inventory in the RCS. However, if the 50 F subcooling cannot be maintained, the HPIAiU shall be immediately reactivated.

HPI#fU flows should also be throttled to prevent violation of the all-ductility temperature (NDT) for the reactor vessel.

This concern can only arise if the fluid temoerature within the reactor vessel is at least 50 F subcooled. A curve 'of the allowable downcomer temperature for a given RCS pressure is provided within the operating guidelines. The downcomber temperature is detemined 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.

2. Without the RC pumps operating, the cold leg RTD's may not provide temperature readings indicative of the actual-RV downcomer tempera-ture as a stagnant pool of water may exist at these locations. The incore thermocouples will provide the best indicator of the down-comer temperature and should be utilized if no RC pumps are available.

In order to account for heat added to the fluid from the core, 150 F must be subtracted from the incore themoccuple readings to reflect the downcomer temperature. This method will result in temperatures which will be lower than the expected downcomer temperature. Thus, use of this methodology assures that NDT will not be & problem.

DATE: 11-20-79 1654 323 PxcE 23

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BABCOCK & WILCOX , , , , ,

NUCLEAR POWtt GENERATION DIVISION 69-1106003-00 TECHNICAL DOCUMENT Figure 1 Break Spectrum-Average System Void Fraction With the RC Pumps Operative and 2 HPI Pumps 100

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BABCOCK & WILCOX . . . .

NUCLEAA POWER GENERATION Ds 43ON 69-1106003-00 TECHNICAL DOCUMENT Figure 2 Pressure Vs Time-Small Breaks with Auxiliary Feedwater 2500 -

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BABCOCK & WILCOX NUQtAA POWER GENERATION DIV133CN 69-1106003-00 TECHNICAL DOCUMENT Figure 3 Pressuri:er Level Vs Time -

Small Breaks with Auxiliary Feedwater 100 i p_ _ q I l 80 -

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DATE: PAGE

SWP-20007 (6-76)

BABCOCK & WILCOX Numsen Nuct Aa powen otNERADON OmslON TECHNICAL DOCUMENT Figure 4 Pressuri:er Level Vs Time for Small Break in Pressuri:er 100 -

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Q 200 400 600 800 1000 Time, see 1654 327 DATE: 11-20-79 PAGE 29

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BABCOCK & WILCOX w ...

maara cowen ceNaATCN QM$loN 69-1106003-00 TECHNICAL DOCUMENT Figure 5 System Pressure Vs Time -

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BABCOCK & WILCOX ..,,

NUCLEAR POWit GENERATCM OW13CN 69-1106003-00 TECHNICAL DOCUMENT Figure 6 Pressuri:er Level Vs Time -

. Class 3 Breaks w/o Auxiliary Feedwater 100 - ___

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1. 1

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BABCOCK & WILCOX Nuuset NyctgAa POWER otNERAfloN DIYl$loN 6S- 6 3-TECHNICAL DOCUMENT INADECUATE CORE COOLING - DESCRIPTION OF PLANT BEHAVIOR

1.0 INTRODUCTION

Following a less-of-coolant accident (LOCA) in which the reactor trips, it is necessary to remove the decay heat from the reactor core to prevent damage. Core heat re= oval is accomplished by supplying cooling water to the core. The water which is available 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 kept 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 beccmes 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 INVENTORY WITH REALTOR COOLANT PIB!PS OPERATING 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. The void fraction of the RCS 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 pu=p 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 AC pumps trip at a time period when the system void fraction is greater than approximately 80%, 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 si:es 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 provide (.. Following the RC pump trip, the small break transient concerns abcut inadequate core cooling will be the same as described in the previws section.

1654 DO -

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BABCOCK & WILCOX wumet NUCL(i R PoWtt otNERAfloN olVl31oM 69-1106003-00 TECHNICAL DOCUMENT 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 eventually become pure steam due to the continued energy addition to the fluid provided by the core decay heat. Under these circumstances, an inadequate core cooling situation will exist. Since the heat removal process under forced circulation is better than the ste1m cooling mode 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 pumps operating.

3.0 LOSS OF RCS INVENTORY WI1HOUT REACTOR COOLANT PUMPS OPERATING Without the RC pumps operating, the cooling of the core is accomplished by keeping the core covered with a steam-water mixture. As the fluid in the core is heated, some of it or all of it may be turned to steam.

If insufficient cooling water is available to maintain the steam-water mixture covering the core, the core exit fluid temperatures will begin .

to deviate from the saturation temperature corresponding to the pressure of tne RCS. One immediate indication that inadequate core cooling may exist in the core is that the temperature of the core exit thermocouples and hot leg RTD's are superheated. At this condition inadequate core cooling is evident as the ' core will be partially uncovered. However, the degree of uncovery is not severe enough to cause core damage. This condition is not expected to occur but is cot, by itself, a cause for extreme action. If the ECCS systems are functioning normally, the temperatures should return to saturation without any actions beyond those outlined for a small break. For incore thermocouple temperature -

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 pump, (c) verify the steam generator level is being maintained at the emergency level (d) if steam generator level is not at 95% of operating range (96 inches indicated on the startup range for raised loop plants), raise level to the 95% level, (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 pressuri:er PORV to reduce RC pressure and reclose PORY when RC pressure falls to 100 psi above the secondary pressure. These actions are directed toward depres- '

suri:ation 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 injection of the HPI system alone is not sufficient to exceed core boil off.

1654 331 DATE: 11-20-79 PAGE 33

. e a e BWP-20007 (6-76)

BABCOCK & WILCOX NUCLEAR POWER otNERAfloN Olvt51oM 68- 6 3-TECHNICAL DOCUMENT If the incere thermocouple indications reach Curve #1 cn Figure 3 in Part I, the peak fuel cladding te=perature has reached approximately 14000F. Above this temperature level there is a potential for cladding Iapture. Also, the :ircaloy 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 frem this level of fuel clad temperature.

For incere the:=occuple te=perature indications at or exceeding Curve

1. 1 on Figure 3 in Part I, the operator should (a) start one RC pump in each loop, (b) depressuri:e the steam generator as rapidly as possible to 400 psig or as far as necessary to achieve a 100 0F decrease in saturation temperature, (c) i= mediately continue the plant cooldown by maintaining a 100 F/hr decrease in secondary saturation terperature 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 pu=ps will provide forced flow core cooling and will reduce the fuel cladding te=peratures. The rapid depressuri:ation of the steam generators will help to depressurize the primary system to the point where the core flooding tanks will actuate. Stopping the depressurizatien at 400 psig (or at a reduction in saturation temperature of 1000F) will =aintain the tube to shell temperature difference within the 100 0F design li=it. The continued cooldown to 150 psig will reduce the primary system pressure to the point where the Low Pressure Inj ection System can supply conling. The opening of the PORY will also help to depressuri:e the primary system. The PORV should be closed when the primary pressure is within 50 psi of the secondary pressure and then shou'.d 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 minimi:e the loss of water from the primary system through the PORV.

If the incore the:moccuple readings reach Curve #2 cn Figure 3 in Part I, 0

the peak cladding temperature is approximately at the 1800 F level. This is c. very serious condition. At this level of clad te=perature, significant amounts of non-cendensible 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 depressuri:e the RCS to a level where the ccre flooding tanks will fully discharge and the LPI system can be actuated thus providing prompt core recovery. The

, operator shculd (a) depressuri:e the steam generators as rapidly as possible down to 50 psia, (b) start the remaining RC pump and (c) open the FORV and leave it open.

1654 332 DATE: 11-20-79 PAGE 34

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BABCOCK & WILCOX .,,

NucteAs town ceNUADoN DM$loN 69-1106003-00 TECHNICAL DOCUMENT 4.0 INADEQUATE CORE CCOLING RESULTING FROM LOSS OF STEN! GENERATCR HEAT SINK For a very small or non-LOCA event, the core decay heat removal is accomplished via the steam generators. If that heat removal is decreased or lost, the natural circulation of fluid within the RCS may be reduced or stopped. The loss of natural circulation for core cooling will eventually boil off the remaining water inventory in the core and lead to inadequate core cooling and elevated core temperature. Indications of loss of steam generator heat sink include (a) a low level in the steam generator with '.ow steam pressure, (b) temperature indicators in hot legs show saturated temperatures, (c) increasing RCS pressure. The operator should try to establish emergency feedwater as quickly as possible and immediately actuate the HPI system to restore natural circulation and RCS heat removal.

If auxiliary feedwater is not available and there is no break in the RCS, the system will repressuri:e and decay heat will be removed by opening the PORY and maximizing HPI addition.

For this plant, system repressurization could result in a loss of the HPI pumps because of the low pressure (< 1800 psig) shutoff head.

For this plant, the combined use of the MU system, startup feedwater pump and the FORV is required to ensure adequate core cooling.

1654 333 PAGE 35 DATE: 11-20-79