ML20003C329
| ML20003C329 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Arkansas Nuclear, Rancho Seco, Crane  |
| Issue date: | 12/11/1980 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20003C324 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-3.A.2.1, TASK-TM 74-1122501, 74-1122501-00, NUDOCS 8102270740 | |
| Download: ML20003C329 (38) | |
Text
B'a?-2000- (6-76; BABCOCK & WILCOX
~sname powie ce~eaariou o.viso TECHNICAL DOCUMENT D!ERGENCY OPERATING SPECIFICATION 74-1122301-00 Doc. ID - Serial No., Revision No.
I for OPERATING GUIDELINES FOR SMALL BREAKS FOR OCONEE 1, 2 AND 3, THREE MILE ISLAND 1 and 2
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LWP-200:5 (6-76; BABCOCK & WILCCX
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RECORO 0F REVISION 3-n:2301 -C i
REV. NO.
CHANGE SECT / PARA.
DESCRIPil0N/CH ANGE AUTHORIZATION 00 Original Issue - Super:edes Preliminary Document 69-1106001-00, Nove:ber, 1979.
In:1udes changes to ensure operation within Brittle Fra::ure b its.
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I DATE:
11-2-80 PAGE :
8 BWP-20006 (6-76)
BABCOCK & WILCOX u-m uvetsas powee ce=reanom suso=
7J-112:501-02 TABLE OF CONTENTS / EFFECTIVE PAGE LIST
$ECTIOld TITLE PAGE 000 NO PART I OPERATING CUIDELINES FOR SMALL BREAKS 1.0 SYMPTOMS AND INDICATIONS (!MMEDIA*E INDICATIONS) 4 74-1122501-02 2.0 IMMEDIATE AC IONS 4
74-112:501-00 i
30 PRICAU* IONS 5
74-11:2301-00 6
74-11:2501-00 7
74-1122501-00 8
74-11:2501-00 4.0 FOLLOWUP ACTIONS 8
74-112:501-00 9
74-1122501-00 10 74-1122501-00 11 74-1122501-00 l
12 74-11:2501-00 13 74-11 2501-00 14 74-1122501-00 APPENDIX A LPI COOLING 15 74-11:2501-00 16 74-1122501-00 17 74-1122501-00 PART II S M L BREAK PHENOMENA - DESCRIPTION OF PLANT BEHAVIOR 10 INTRODUCTION 18 74-11:2501-00 2.0 IMPACT'0F RC PUMP OPERATION ON A SMALL 18 74-1122501-00 LOCA 19 74-1122501-00 3.0 5 % L BREAKS WITH AUXILIARY FEEDWATER 19 74-1122501-00 20 74-1122501-00 21 74-1122501-00 22 74-112:501-00 I
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4.0 SMALL BREAKS WITHOUT AUXILIARY
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22 74-11:2501-00 FEEDWATER 23 74-1122501-00 24 74-1122501-00 l
TRANSIEE S WITH INITIAL RESPONSE 5.0 SIMILAR TO A SMALL BREAK 24 74-1122501-00 25 74-112:501-00 6.0 TRANSIENTS THAT MICHT INITIATE A 25 74-1122501-00 26 74-1122501-00 LOCA l
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I PAGE 3 l-DATE:
12-3-80 l
B WP-2000e M-76)
BABCOCK & WILCOX ms-s t e Nu;Llas powre GtNteatQm Drvisom TABLE OF CONTENTS / EFFECTIVE PAGE LIST 7 - n2: sci-oc
$ECTION TITLE PAGE 000. No.
7.0 EPI THROT :.!NG 26 74-11:2501-00 27 74-112:501-00 28 74-1122501-00 FIGURE 1 RC PRESSURE /TD'.PEPa!URE LIMITS 29 74-112:501-00 FIGURE 2 PRESSURE 75 TIME-SMALL BREAKS W:*H AUXILIARY FEEDWATER 30 74-112:501-30 FIGURE 3 PRESSURIZER LEVEL. VS TIME-SMALL BREAKS WI*H AUX!LIARY FEEDWATER 31 74-112:501-00 FIGURE 4 PRESSURIZER LEVEL VS TIME FOR SALI.
BREAK IN PRESSURIZER 32 74-112:501-00 FIGURE $
SYSTEM PRESSURE VS TIME-SALL BREAKS 33 74-11:2501-00 W/0 AUXII.!ARY FEEDWA*F.R FIGURE 6 PRESSURI*ER LEVEL VS *IME-CLASS 3 BREAKS W/0 AUXILIARY FEEDWATER 34 74-112:501-00 APPENDIX A INADEQUATE CORE COOLING - DESCRIPTION OF PLANT BEHAVIOR 35 74-11:2501-00 36 74-112:501-00 37 74-1122501-00 38 74-1122501-00 t
PAGE 3-1 DATE:
12-3-80 I
32N?-20007 (6-76; BABCOCK & NVILCOX
- 8 I' sew;ttaa powie otNtaattom 0,vto%
7'-11:1501-3; TECHNICAL DOCUMENT PART : - OPERA!!NO C"!DE* :NES FOR SMAL*. 3REAl*S 1.0 P.?.?*0MS AND INDICATIONS (IMMEDIATE INDICATIONS) 1.1 Excessive reactor coolant sys:e= (RCS) =akeup*
12 Decreasing ROS pressure 1.3 Reactor trip 14 Decreasing pressurizer level
- 1.5 ESFAS actua: ion
- 16 Low =akeup tank level
- 1.7 Additional criteria during heatup and cooldown*
1 7.1 RCS temperature increasing, minimu= le:down and pressuri:er level decreasing.
- ith a cooldown of < 1700F/hr and cannot =aintain level in =akeup 1.7.2
=
tank.
2.0 IMMEDIA~E AC IONS 2.1 If the ESTAS has been initiated automatically because of low RC pressure, immediately 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 Balance-high-pressure injection (HPI) flow between injection lines when HP1 is' initiated.
l
'4 Verify : hat appropriate once-through steam generacor (OTSG) level l
1s maintained by feedvater control (low level limit vi:5 RC pu=ps operating, emergency level without RC pu=ps operating).
2.5 Monitor sys:e= pressure and tempera:ure. If satura:ed conditions occur, initiate HPI.
2.6 If ESTAS has been bypassed due to heatup or cocidown, ini:ia:e i
safacy in4ection.
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l
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l CAUTION:
If 50 F subcooling criteria is met, thro::le EPI flow per Step 3.4 If RCS is not 500F subcooled, continue full safety injee: ion until 500F subcooling is attained.
- May not occur on all small breaks.
l DATE:
12-3-80 PAGE 4
o
a Ek*NF-2000' (6 'i, BABCOCK & WILCOX w;' t A8 'owt8 Gl= Hatch civisso=
TECHNICAL DOCUMENT 7 -112:501-o' 3.0 PRICAL*TIONS 3.1 If the ESFAS has been ini:iated on low RC presure, :er=ina: ion of RC pu=p operation takes precedence over all other i= mediate actions.
NOTE: If ESFAS has been a :uated on high R$ pressure, then monitor RC pressure and trip RC pc=ps once pressure decreases below the ESFAS low pressure se: point.
3.2 If ESFAS has been iniziated, the RC pu=p's tripped, and the RCS deter =ined to be at least 500F subcooled, the opera:or should establish as quickly as possible if the cause for :he depressuri-
- stion is due :o either a LOCA or non-LOCA (overcooling) event.
Proceed to Step 4.4 for non-LOCA events.
- 'ith no RCPs running, the degree of subcooling is deter =ined NOTE:
=
by averaging the five highest incore thermocouple te=perature readings.
3.3 If '.he HPI syste= 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 CPM
- in each line and the situation has been stable for 20 =inu:es.
or 2.
All hot and cold leg temperatures are at least 50 F below the saturation temperature for the existing RCS pressure
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and l
l the action is necessary to prevent the indicated pressurizer level from going off-scale high.
l NOTE: If 500F subcooling cannot be maintained, the HPI shall be reac:ivated.
l NOTE: The degree of subcooling beyond 500F and the length of time HFI is in operation shall be limited by the pressure /
temperature considerations for :he vessel integrity (see l
Section 3.4).
I i
l
I I
DATE: 12-3-80 PAGE 5
EWNF-20007 (6-761 BABCOCK & %VILCOX "U"8 !'
mu?'las sowta otwisatiCN civtsio%
TECHNICAL DOCUMENT
- -:1::50:-oc-3.4 Pressure / Temperature considerations for vessel integri:y are dependent on whether or no: RC pumps are operating as follows :
3.4.1 If RC oumos are coerating, and the reactor coolant is > 500F subcooled, the reae:or vessel downco=er pressure /tempeiature (?-T) ce=bination shall be kept within the nor=al NDT limits of technical specifications.
NOTI: With one or more RC pumps operating use any cold leg RTD as an indication of reactor vessel downcocer t em pe ra ture.
3.4.2 If 25[ RC pumps are operating, the RC pressure / temperature combination i
shall be kept within the no forced flow region of Figure 1.
NOTE: With }$! RC pumps operating, :he RC temperature shall be de: ermined by averagining the five highest incore thermocouple temperature readings.
3.4.2.1 When :he reactor coolant is 1 500F subcocled, con:inually reduce the HPI flow rate to main:ain the P/T limits of Figure 1.
NOTI: Maintaining the reactor coolan: 500F subcooled takes precedence i
over the Brittle Fracture Limit of Figure 1.
3.4.2.2 As soon as RC pressure / temperature is below the maximum limit of the DERS, start the DERS and paintain the RC pressure / temperature within the limits of the DERS.
NOTE: Until the HPI flow is ter=inated, the temperature must be, kept within the NO Forced Flow regivn of Figure 1.
35 Pressuricer level may be increasing due to RCS reaching saturated condi: ions or a break on top of the pressuricer.
3.6 If high activity is detec:ed in a stea= generator, isolate the leaking generator. 1: is recommended tha: both steam generators not be isolated.
3.7 Other indications which can confirm the existence of a LOCA:
3.7.1 RC drain tank (quench tank) pressure (rupture disk may be blown).
3 7.2 Increasing reac:or building sump level.
3.7.3 Increasing reacter building temperature.
i I
DATE:
IAC 12-3-80 6
j.
. ~,.
3WN?-20007 (6-76' BABCOCK & WILCCX hw;Llas powte Gi%taa'!Oh o'v$Ch TECHNICAL DOCUMENT 7 -n2:5x -c; 3
7.-
Increasing reac:or building pressure.
3.7.5 Increasing radiation monitor readings inside containment.
3 7.6 Reactor coolant system te=perature beco=ing sa:ura:ed relative to the RCS pressure.
i 3 7.7 Ho: leg tempera:ure equals or ex:eeds pressuri:er :emperature.
3.7.8 Increase in :he excore neutron detector indications.
NOTE: In conjune: ion with the indica: ions in 3.10.1, this could be an indication of inadequate core cooling.
3.8 EPI cooling requiremen:s could deplete the borated water storage tank, and initiation of LPI flow from the reactor building su=p 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 pressurizer level).
3.10 Maintain a :emperature versus time plot and a corresponding temperature pressure plot on a satura: ion diagram..Using ho: leg RTD's and highest incore ther=occuple reading, these plots will make it possible to.: rack the plant's condi: ion through plant cooldown.
3.10.1 If either of the following indications of inadequate core cooling exist, go to Section 4.5.
1.
Hot 1,eg 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 is decreasing along the satura-tion curve then subcooled conditions will be established. This will be indicated by primary system pressure no longer following the saturation curve, as primary system temperature decreases. When this occurs, primary system pressure should be controlled by adjusting EPI flow, to maintain 500F subcooling. The degree of subcooling beyond 500F shall be controlled within the limits defined in Section 3.4.
3.11 Component cooling water (CCW) and seal injec: ion should be maintained to the RC pumps to insure continued service or the ability to restart the pumps at a later time.
DATE:
12-3-80 pAGE 7
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4 l
3*.*N7-20007 (6-76 ;
~uai..O.o.K. & WILC O X 6ABC C i
oi~.oe~e-re~
TECHNICAL DOCUMENT 7 -1:::501-o:
i 3.11.1
,:=al limits and precau: ions apply for RC pun; opera:icn.
3 11.2 If :he RC punps are tripped for any reason, seal inje:: ion should be maintained tc ensure long tern seal integrity.
l 4.0 FOLLO*JL*P ACTIONS P
a.1 Identification and Early Control 4.1.1 If HPI has initia:ed because of low pressure, control HPI in accordance vi:h Steps 3.3 & 3.4 4.1.2 If both HPI trains have not actua:ed on ESFAS signal, star: second HPI train if possible. Balance KPI flows.
4 1.3 If R7 pressure de:reases continuously, verify that core flood :anks (CFTs) and Icw pressure injection (LPI) have ac:ua:ed as needed, and balance LPl.
4.1.4 If cause for cocidown/depressurization is determined to be due to a non-LOCA overcooling event and the ROS is a: leas: 500F subcooled then proceed to Section 4.4.
4.1.5 A: esp: to locate and isolate leak if possible. Latdown was isolated in Step 2.2.
0: hor isola:able leaka are PORY (close block valve) and between valves in spray line (close spray and block valve).
i 4.1.6 Determine availabili:y of reactor coolan: pumps' (RO?s) and nain and auxiliary feedwater systems. If feedwa:er is not available, go to 4.2.
If feedwater is available, go to 4.3.
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4.2 Actions if Feedwater is not Available
4.2.1 Throughou
the following s:eps maintain maxinu= HFI flow per S:ep 3.4 and res: ore feedwater as soon as possible.
4.2 2-If the subcooling margin is adequate and RCPs are operating, go to one pump. If RCPs are not operating, go to Step 4.2.5 below.
4.2.3
_ If RCS pressure increases, open FORV and leave open.
f NOTE: If the PORY cannot be actuated, the safeties will relieve pressure.
DATE:
12-3-80.
PAGE b
__ __. _.. _ ___ __ _.m f
3RN7-20007 (6-76) i GABCOCK & WILCOX
}
hu '.taa oowte otmteatC% Divisiom 7 -11225cl-::
TECHNICAL DOCUMENT i
4.2.4 When fee: water is recovered, restore OTSO levels in a controlled l
manner. Close PORV or block valve, if possible. Proceed to Step l
4.3.2.
4.2.5 If no RCPs are operating, open PORV, maintain HPI flow.
NOTE: If the FORY cannot be actuated, the safeties will relieve pressure.
- 4. 2. 6 When feedwater flow is restored, raise OTSO levels to 952 on the operate range, close FORV or block valve, if possible.
NOTE: OTSO level should be monitored periodically during the fill process. Levels > 95% on the operating range must be avoided to preclude feedwater carryover to the steamlines.
4.2.7 Verify natural circulation in the RCS by observing:
i 4.2.7.1 Cold leg temperature is saturatien temperature of secondary side pressure within approximately 5 minutes.
4 2.7.2 Primary AT (THOT-TCOLD) becomes constant.
i 1
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 one RCP running per loop (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 1000F per hour to achieve an RC pressure of 250 psig. Refer to Precaution.3.10 for i
development of temperature and pressure plots. Isolate core flood tanks when 500F subcooling is attained and RC pressure is less than 700 PSIC. Go into LPI cooling per Appendix A.
t 4.3.4 If RCPs are not operating :
i 4.3.4.1 Establish and control 0TSG 1evel to 95% on the operate range. Verify the conditions in Step 4.2.7.
NOTE: OTSG levels greater than 952 on the operating range must be avoided to preclude feedwater carry-over into the steamlines.
I PAGE 9
DATE:
12 2-80
s'.5?-20007 (6-76; B ABCOCK & Wl'.COX
.m gntaa somit otweeabo% olv'5.ON TECHNICAL DOCUMENT 7;-11:2501-00 4.3.4.2 if R; pressure is decreasing, wait until it stabili:es or begins in:reasing.
If it begins increasing, go to Step 4. 3.4.e.
4.3.4.3 Proceec with a controlled cooldewn at 100oF/hr by controlling stea=
genera tor secondary side pressure. Monitor RC pressures and te=peratures 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 syste= te:peratures 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 te=perature for the existing RCS pressure.
and B) The hot and cold le3 temperatures are decreasing in response to stea= generator secondary te=perature decrease.
If these conditions are satisfied and remain satisfied, continue cooldown to achieve an RCS te=perature (cold leg) of 2800F, and proceed to Step A.1 of Appendix A.
NOTE: If the conditions above are met below 700 PSIG, the core flood tanks should be isolated.
i 4.3.4.3.3 Start a reactor coolant pump, if the prieary sys tem is 500F sub-cooled in both hot and cold legs and pri=ary syste= pressure is above.'50 PSIG. If the 500F subcooling =argin is ever lost, i==ediately trip RC Ps.
If forced circulation is achieved, proceed I
to Step 4.3.
4.3.4.3.4 If RC pressure stops decreasing and the conditions of 4.3.4.3.2 are not met or cease to be =et or if RC pressure begins to increase, then proceed to Step 4.3.4.4 below.
4.3.4.4 Restore R:P 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 pre..iure is increasing, starting a pu=p is per=issible at RC pressure greater than 1600 PSIC.
FACE DATE:
12 3-80 10
Si'N?-20007 (6-76 ;
BABCOCK & WILCOX
"'"8" uceueewnoe nco smse TECHNICAL DOCUMENT 74-11:2501-00 4.3.4.4.2 If reactor coolant syste= pressure exceeds stea= generator second-ary pressure by 600 PSIG or = ore "bu=p" one reactor coolant pu=p for a period of approxi=ately 10 seconds (preferably in operable stea= generator loop). Allow reactor coolant sys:e= pressure to stabili:e. Continue cooldown. If reactor coolant syste: pressure again exceeds secondary pressure by 600 psi, wait at least 15
=inutes and repeat the pu=p "bu=p".
Bu=p al:ernate pu=ps so tha:
no pu=p is bu= ped = ore :han or.ce in an hour. This =ay be repea:ed, with an interval of 15 =inutes, up to 5 ti=es.
Af ter the fif th "bu=p", allow the reactor coolant pu=p to con:inue in operation.
4.3.4.4.3 If pressure has stabilized for greater than one hour, secondary pressure is less than 100 PSIG and pri=ary pressure is grea:e than 250 PSIG, bu=p a pu=p, wai: 30 =inutes, and star an alternate pu=p.
4.3.4.5 If forced flow is established, go to Step 4.3.3.
4.3.4.6 If a reactor coolant pu=p cannot be operated and reactor coolant syste= pressure reaches 2300 PSIC, open pressurizer PORV to reduce reactor coolant syste= pressure. Reclose PORV when RCS pressure f alls to 100 psi above the secondary pressure. Repeat if neces-sary. If PORV is not operable, pr essuri:er 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 = ode until m RC pu=p is started or stea:
generator cooling is established as indicated by establishing con-ditions described in 4.3.4.3.1 or 4.3.4.3.2.
When :his occurs, proceed as directed in those stepi.
Go to Step 4.3.2 if forced flow is established.
4.4 Non4LOCA Overcooling Transient with Feedwater Available
- 4. 4.1 l==ediately restart a RC pu=p in each loop if the RCS' is 500F subcooled.
4.4.2 Control stea= pressure via turbine bypass or at=ospheric du=p valves to stabilize or control plant heatup.
NOTE: Considerable HPI =ay have been added to the RCS.
There-fore, to prevent RCS fro = going solid, the above action
=ay be necessary.
4.4.3 As long as _ the RCS is =aintained 500F subcooled, throttle HPI/MU and letdown flow to =aintain pressurizer level at N 100 inches.
PAGE gi DATE:
12-3-80
EWNF-20007 (6-76)
BABCOCK & WILCOX meUOf A8 Powf a otNft Ahoh oW!$3oN TECHNICAL DOCUMENT 7 -::22501-00 4.4.4 Using turbine bypass valves and feedwater system, con:rol stea:
generators as needed to limi: plant hea:up until RC pressure control can be re-established with the pressurizer.
NOTE: Gold RCS water may have been added to the pressuriser; therefore, a period of time may elapse before nor=al RC pressure control can be established with the pressuri:er heaters.
4.4.5 Once pressure control is re established, use nor=al heatup/cooldown procedure to establish desired plant conditions.
4.5 Actions for inadecuate Core Cooling 4.5.1 Immediate steps for inadequate core cooling NOTE: If RC pumps are running, do not trip pumps. This supercedes instrue:1ons in Section 2.1.
4.5.1.1 Verify EPI /LPI systems are functioning properly with maximum flow.
Start makeup pu=p(s), if possible, to increase injection flow.
4.5.1.2 Verify steam generator level is being con: rolled at 95% on operate range.
CAUTION: Reference leg boiling could give false level indication.
4.5.1.3 Depressurize operative steam generator (s) to establish a 1000F/hr decrease in secondary saturation temperature.
~
4.5.1.4 Ensure core flood tank isolation valves are open.
l 4.5.1.5 If reactor coolant system pressure increases to 2300 PSIG open pressuri:er PORV to reduce reactor coolant system pressure.
Re-I close PORV when RCS f alls to 100 PSIG above :he secondary pressure.
l Repeat if necessary. If PORV is not operable, pressurizer safety l
valves will relieve pressure.
l 4.5.1.6 Proceed immediately to 4.5.2.
- 4. 5. 2 When the indicated incore thermocouple temperatures or hot leg RTD temperatures are superheated for the existing RCS pressure, opera-tor ac: ion shall be based on conditions determined from Figure 3, by a sample of the highest incore thermocouple temperature readings I
to determine the core exit thermocouple temperature.
i l
NOTE: More than one :hermocouple temperature reading should be used (for example use the average of 5).
4.5.3 When the incore thermocouple temperature has been determined per t
l Section 4.5.2, go to the see: ion indicated below.
DATE:
12-3-S0 pAGE 12 l
l l
Sk':?-20007 (6-76)
BABCOCK & VVILCOX "U""'
Nyctaa powie otNitAhok D'vlSiON TECHNICAL DOCUMENT 72-112:5:1-0c Incore Ther=ocouple Te:cerature Section Incore Tc < Saturation 4.1.6 Curve 1 $, Incore Tc < Curve 2 Figure 3 4.5.4 Incore Tc 3, Curve 2 Figure 3 4.5.5 NOTE: The incere thermocouple temperature readings shall be continuously monitored. If the temperature is between saturation and Curve 1 Figure 3, only the preceeding actions will be taken un:11 the indicated incore ther=o-couple temperatures return to saturation te=perature for the existing RCS pressure or the tempera;ure increases to Curve 1 Figure 3.
4.5.4 Ac: ions for curve 13, incore Tc < Curve 2 Figure 3.
4.5.4.1 If RC pumps are not operating, start one pu=p per loop (if po s sible). This instruction supersedes previous instrue: ions to trip RC pumps.
NOTE: Do not bypass normal interlocks.
4.5.4.2 Depressuri:er operative stea= gecerator(s) as rapidly as possible to 400 PSIG or as far as necessary to achieve a 1000F decrease in secondary saturation te=perature.
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 depressurization was no: possible, open PORV and leave open.
4.5.4.4 lamediately continue plant cooldown by maintaining 1000F/ hr.
Decrease in secondary saturation temperature :o achieve 150 PSIG RCS pre s sur e.
CAUTION: If auxiliary feed pump is supplied by main steam, do not decrease pressure below that pressure necessary for auxiliary feed pump operation.
4.5.4.5 If the average incore thermocouple 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 3, Curve 2 Figure 3 pAGE 13 DATE:
12-3-80
o BWN?-20007 (6-76 BABCOCK & WILCOX
- .v:;taa Powta GE=taafiom D'v:54em TECHNICAL DOCUMENT 73-11:2531-oc 4.5.5.1 If possible, s: art all RC pu=ps.
Starting in:erlocks should be defeated if necessary.
4 NOTE: In order to minimize the possibility of a fire due to bypassing some interlocks, the following precautions should be observed:
- 1) Do not defea: the overload trip circuit and
- 2) If CCW is not restored to the motor wi:hin 30 =inutes, trip the RC pump.
It shculd be recogni:ed that s:ar:ing the RC pumps withou: cooling and/or inje::1on water will probably fail the pu=p seals and =ay cause the pu=p shaf: to break.
However, some core cooling will be provided prior to destruction of the pu=p.
Breakage of :he pu=p shaf: will no:
cause consequential da: age outside of the pump.
4.5.5.2 Depressurize the operative steam generator (s) as quickly as possible to atmospheric pressure.
CA'JTION:
If auxiliary feed pump is supplied by main steam, do not decrease pressure below that pressure necessary for auxiliary feed pump operation.
4.5.5.3 Open the pressurizer PORY and leave open.
NOTE: The RCS will depressuri:e and the LPI system should restore core cooling.
4.5.5.4 When incore thermocouple :emperatures return to the saturation te=perature for the existing RCS pressure; and the LPI syste= is delivering flow, proceed as follows:
4.5.5.4.1 Close the pressurizer PORV; reopen if RCS pressure increases above 150 PSIG.
4.5.5.4.2 Decrease to two (2) RC pump operation (one per loop).
4.5.5.4.3 Isolate the core flood tanks.
4.5.5.4.4 Maintain s:eam generator pressure at atmospheric or as low as possible if maintaining auxiliary feed pump in operation off of =ain stems.
l' 4.5.5.4.5 Control EPI per 3.3.
4.5.5.4.6 Monitor BL'ST level as lo-lo level li=its are approached, align LPI system for suction from R3 sump. Close the LPI BWST suction valves.
NOTE: If HPI is required per 3.3, align LPI and EPI in piggyback mode. Close HPI suction valves to BWST.
- 4. 5. 5. 4. 7 Go to Appendix "A".
DATS:
12-3-80 pAGE 14
BUNP-20007 (6-76)
BABCOCK & WILCOX ugataa powta otmtaatio civ:54o 7,_.
5 3 -00' TECHNICAL DOCUMENT APPENDIX A - LPI COOLING a.1 Deter =ine if primary coolant is a: least 500F subcooled.
If not, go to Step A.3.
A.l.1 Star: LPI pu=ps.
If both pumps are operable, go to Step A.2.
For one LP pump operable =ain:ain OTSO cooling and proceed as follows.
The operable LPI pu=p will be used to =aintain system inventory.
A.l.2 Obtain pri=ary syste= conditions of i 2800F and 1250 PSIG.
A.l.3 Align the discharge of the operable LPI pu=p to the suctions of the HPI pu=ps and take suction fro = the BWST.
If the SWST is a: the low level alar =, align LP! suction from the R3 su=p and shut sue: ion from BWST.
A.l.4 Star: the operable LPI pu=p specified above. The HPI-LPI syste=s will now be in " piggy back" and EP! flow is maintaining syste=
pressure.
A.1.5 Go 'to single RO pu=p operation.
A. l. 6 -
When the second LPI pump is available, align it in the decay heat
= ode and co==ence decay heat re= oval. (Decay heat syste: flow greater than 1000 GPM)*.
Secure re=aining RC pu=p when decay hea:
recoval is established.
C A"TIJN: Verify that adequate NPSH exists for the decay heat pu=p in the DH re= oval mode.
If inadequate, trans f er to LPI = ode.
A.l.7 Reduce reactor coolant pressure to 150 PSIG by throttling HPI flow.
Control RC temperature using the decay heat systa= cooler bypass to
=aintain syste= pressure at least 50 psi above saturation pressure, to assure : hat NPSH require =ents for the decay heat pu=p are main-tained.
A.l.8 Secure the HPI pu=p and shift the LPI pump supplying it to the LP!
injection = ode.
A.l.9 Reduce reactor coolant temperature to 1000F by controlling the decay heat syste= cooler bypass.
NOTE: If one of the LPI/ decay heat pumps is lost, return to OTSG cooling using natural circulation or one reactor coolant pu=p.
Go to A. l. l.
- For Arkansas Nuclear 1 use 2630 GPM.
DATE:
12-3-80 PAGE 15
'BL*NF-20007 (6-76; BABCOCK & WILCOX NU*'8E8 wu Liat *owie ofwteanom oiviseow TECHNICAL DOCUMENT 7'-1:22501-oe A.
Cooldown on Two LPI Pucos A.O.1 Maintain RCS pressure at < 250 PSIG and reduce RCS te=perature to
~
1 280 F.
A.2.2 Align one LPI pu=p in the decay heat re= oval mode.
A.2.3 Secure one RC pu=p if two are operating.
A.2.4 Start the decay heat pu=p in the decay heat removal = ode, and when i
decay hea: syste= flow is grea:er :han 1000 GPM*, secure the run-ning RC pu=p.
A.2.5 Reduce RC pressure to 150 PSIG by throttling HPI flow. Control RC temperature to =aintain at least 50 psi margin to saturation pr e s sur e.
A.2.6 Start the second LPI pump in the LPI injection mode. Secure HPI pump.
A.2.7 Shif t LPI suction f rom the BWST to the reactor building su=p when lo-lo level li=its are approached.
A.2.8 Reduce reactor coolant temperature to 1000F by con: rolling the decay heat system cooler bypass.
NOTE: If one of the LPI/ decay heat pumps is lost, return :o OTSG cooling using natural circulation or one RC pu=p.
Go to A.l.l.
A.3 Cool Down RC System at Saturation A.3.1 Maintain RC pressure at < 250 PSIC.
A.3.2 Align one LPI pu=p to suction of the HPI pu=ps and the suction to the reactor building sump.
(Shu: BWST suction valve for this pu=p.)
l A.3.2 When the BWST level reaches the lo-lo level li=its, star: the LPI pump and shu: the RP! pump suction from the BWST.
A.
When primary system temperature becomes subcooled b;* at least 500F, go to A.1 1.
A.4 Cooldown without Reactor Coolant Pumps i
[.
A.4.1 RCS initial conditions are: pressure 150 psi, te=perature at satura-tion.
A. 4. 2 ~
Align low pressure injection system for suction from reactor building sump ' and place into service.
- For Ar*eansas Nuclear 1 use 2630 GPM.
DATE:
12 2 80 PAGE 16 i
BWF-20007 (6-76)
BABCOCK & WILCOX "d" 8 E 8
%ctitas sowie Ggwgaat:Ch Div?SiON TECHNICAL DOCUMENT 7;- n::3a-ce A.4.3 Balance LPI injec: ion and con:rol RC :e:perature vi:h decay heat coolet,.
A.4.4
!solate core flood tanks.
A.4.5 Go :o step A.l.1 and follow the procedure given there, ignoring the instrue ions relating to RC pu=p operation.
J 4
1 I
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4 i
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I l
i DATE:
12-3-80 PAGE 17 I
t L
EWN?-20007 (6-76)
B ABCOCK & WILCOX Wa* 8 t t
%U:Llat Powll otNI# Af ton Divt5loh TECHNICAL DOCUMENT 74-11:2501-e:
PART II: SMA1.L BREAK PHINCENA - DESCRIPTION OF PLA',T BERAGOR 1.0 I!.iRODUC~ ION A loss-of-coolant accident is a condition in which liquid inventory is lost from the reactor coolan: system. Due :o :he loss of = ass fro: :he reactor coolant systa=, the most significant sho r - te r:
sy=pte: of a loss-of-coolant acciden is an uncontrolled redue: ion in the reactor coolant syste: pressure. The reactor protection system is designed to trip the reactor on low pressure. This should occur before the reactor coolant syste reaches saturation condi-tions. The existence of saturated conditions within the reactor system is the principal longer-ter= indication of a LOCA and re-quires special consideration in the development of opera:ing procedures.
Following a reactor trip, it is necessary to remove decay heat fro =
the reactor core to prevent da= age.
However, so long as the reac:or core is kept covered with cooling water, core damage will be avoid-ed.
The ECCS sys: ems are designed to respond automatically to low reactor coolant pressure conditions and take the initial actions to protect the reactor core. They are sized to provide sufficient water to keep the reactor core covered even with a single f ailure in the ECCS systems. Subsequen: operator actions are required ultimate-ly to place the plant in a long-ters cooling mode. The overall objec:ive of the automatic emergency core cooling system and the followup operator actions is to keep :he reactor core cool.
A detailed discussion of the small break LOCA phenomenalogy is pre-sented in this section. This discussion represents Part II of :he 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 pres-surizer level histories for various combinations of parameters are presented in order to indicate the wide range of sys:e= 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 Figute 1.
The maximu= void fraction that the syste= evolves to, and the time DATE:
12-3-80 PAGE 18
EWNP-20007 (6-76; BABCOCK & WILCOX
%vut ta NW3tas powen otNtaatio% Divisio%
TECHNICAL DOCUMENT 7 -::2250:-c:
it occurs, is dependent on the break size and loca: ion.
Continued RC pu=p operation, even at high syste: void fractions, will provide sufficient core flow to keep cladding temperatures within a few de-grees of the sa:urated fluid temperature.
Since the RCS can evolve to a high void fraction for certain s=all breahe tath the AC pumps on, a RC pump trip by any means (i.e., loss of off/ite power, equipment failure, etc.) at a high void fraction durint the small break transien: =ay lead to inadequate core cool-ing. That is, if the RC pu=ps trip at a ti=e period when the sys:e:
void fraction is greater than approximately 70*, a core heatup will occur because the amount of water lef t in the RCS would no: be sufficient to keep the core covered. The cladding te=perature 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 te=peratures 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 coc'.ng is provided. Fciloving the RC pump trip, the small break transient will evolve as described in the subsequent sections.
3.0 SMA1.L BREAKS WI~E AUXILIARY FEEDWATER There are four basic classes of break response for small breaks with auxiliary feedwater. These are:
1.
LOCA large enough to depressurize the reactor coolant system.
2.
LOCA which.Labilizes at approximately secondary side pressure.
3.
LOCA which may repressurize in a saturated condition.
4.
Scall LOCA which stabilizes at a primary system pressure i
greater than secondary syste= pressure.
The system transients for these breaks are depicted in Figure 2.
3.1 LOCA Large Enouch to Deoressurize Reactor Coolant Syste:
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 pres-
~
surize the system to a stable low pressure. ECCS injection easily exceeds core boil-off and ensures core cooling. Curves 1 and 2 of I
Figure 3 show the pressurizer level transient. Rapidly falling pressure causes the ho: legs to saturate quickly. Cold leg tempera-ture 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 l
i i
DATE:
12-3-80 PAGE 19 I
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TECHNICAL DOCUMENT 7-n 52-2 of the stea: generators, they are essentially unaf f e::ed by the availabili:y of auxiliary feedva:er. Upon receipt of a low pressure ESFAS signal, the operator mus: : rip all RC p=ps and verify tha:
all ESyAS actions have been co=pleted. he operator =ust als; balance EPI flows such tha: flow is available through all W.?'
inje:: ion no::les even if only one E?! is available.
~he operator should also balan:e !.l? flows, should :he syste= be at:uated, ::
ensure flov through both lines.
~he opera:or needs to take no further a:: ions to b. *ng :he syste= to a safe shutdown ondi:1:n.
Rapid depressuri:ation e~
- ne stea: genera: ors would on' y ac: to accelerate RCS depressurica: ion.
- is, however, no: ne.:e s s ary.
Restar:ing of :he RC pu=ps is no: desirable for :his class of break.
14:g-ter: cc oling vill require :he operator :o shif: the L?* pu=p sue: ion to ;he rea::or building su=p.
3.2 LJCA which Stabilizes at Accroxi=ately Se:ondarv Side Pressure Curve 3 of Figure 2 shows the pressure transien: for a break vnich is too small in co=binatice wi:h :he operating H"I to depressurize the ROS.
- he stea: generators are, therefore, necessary to re=ove a portion of core decay hea:. Although :he sys:e pressure vill ini-tially stabili:e near the secondary side pressure, RCS pressure =ay eventually begin falling as the decay hea: level decreases. Curve 3 of Figure 3 shows pressurizer level behavirr.
he ho: leg tempera-ture qui:kly equalizes to :M saturated :e:perature of the secondary side and controls primary syste pressure at saturs: ion.
~he cold leg te=perature may re=ain slightly subcooled. If :he EP: refills and repressurizes the RCS, the ho: legs can beco=e subcooled.
~he 1:=ediate operator action is to trip the RC pu=ps upon receipt of
- he low pressure ESFAS signal and then verify ESTAS functions.
~he operator mus: then balance HP! in order to ensure flev :hrough each high pressure injection line.
t i
Follevup action by the operator is to raise the e=ergen:y feedwater l
level to 951 on the operating range and che:k for established natural circula:icn.
"his is done by gradually depressuri:ing the I
seca: generators. If this test f ails, in:er=ittent bu= ping of a RC pu=p should be perfor=ed as soon as one is available. Con:inued depressurination of :he s:ca: generators vi:5 na: ural circula: ion leads :o cooling and depressuri a: ion of the RCS.
- he opera:or's goal is to depressurize the RCS to a pressure : hat enables the ECOS to exceed-core boil-off, possibly refill the RCS, and to ultica:ely
~'
es:ablish long-ter= cooling.
l 3.3 LOCA which sav Repressuri:e in a Saturated Condition i
l Curve !. of Figure 2 shows the behavior of a s=all break : hat is :co s=all, in co=bination with the HPI, to depressurize :he pri=ary syste:. Although s:ca generator feedwater is available, the loss DATE:
PAGE 3;
37 3_33 i
s I
375?-10007 (6-76; BABCOCK & WILCOX wg stas powtB otNES ATCh Dm5 TON TECHNICAL DOCUMENT 7e-n :2501-oc of pri=ary syste: coolant and the resultan R05 voiding will eventually lead to interruption of natural circulation. This is followef by gradual repressuri:ation of the pri=ary system.
I: is possible that the primary syste could repressurize as high as the pressurizer safety valve se: point before the pressure stabili:es.
This is shown by the dashed line in Curve 4 Once enough iuven:ory has been lost from :he primary sys t= :o allow direct stea: conden-sation in the regions of the s ea generators contacting secondary side coolant, the pri=ary syste= is forced to depressuri:e to the saturation pressure of the secondary side.
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 sufficien: to replace the inven-tory lost to boiling in the core, and condensation in the stea:
generators re= oves decay heat energy. The RCS is in a stable thermal condition and it will remain there until the operator takes fur:her 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 repressurizat.on pha:e. The dashed line indicates the level behavior if pressure is forced up to the pressurizer safety valve se: point. During this transient, hot leg te=perature will rapidly approach saturation with the initial syste:
depressurisation, and it will re=ain saturated during the whole transient. Cold leg temperature will approach saturatian as circula-tion is lost, but may re=ain slightly subcooled during the repres-surization phase of the transient. Later RCS depressurization could cause the cold leg te=peratures to reach saturation. Subsequent refilling of the primary system by the HP1 might cause temporary interruption of steam condensation in the stea= generator as the primary side level rises above the secondary side level. If the depressurization capability of the break and the HP! is insuf ficient to offset decay heat, the primary system will once more repressur-ize. 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 stea= generator. This cyclic behavior will s:op once 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 ESTAS functions. The operator should then balance HPI flow.
Following that, he should raise the steam genera:or level to 95% of the operating 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 estab-lish long-term cooling. If the system fails to go into natural circulation, he thould open the PORV long enough to bring and hold the RCS near th4 secondary side pressure. Once natural circulation is established or a RC pump can be bumped, he will be able to con-tinue depressari:ing the RCS with the stea: generators and establish long-term cocling.
PAG 2 21 DATE:
12-3-80
f BkT?-20007 (6-76)
B ACCOCK & WILCC'X uv.. i.
,vmgaa Powie otNteA7loN Divi 5loN i
' a ' ^' " 501-0 0 TECHNICAL DOCUMENT
~
3.4 S=all LOCA which Stabilizes at P > Psee 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 emp:1ed. The pri=ary syste: re-mains subcooled and natural circulation to the steam generator recoves core decay heat.
The pressurizer never e=pties and continues to con-
- rol pri=ary syste pressure. The operator needs to trip the RC pumps and ensure that ESFAS actions have occurred. Throttling of HPI is per=itted only af ter RCS subcooling of 500F has oeen established, the i
pressurizer has refilled, and natural or forced circulation has been
(
verified. A restart of the RC pumps under these conditions is desirable for plan: control.
l 3.5 Small Breaks in Pressurizer The system pressure transient for & small break in the pressurizer will behave in a manner similar to that previously 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 transient, the pressurizer may contain a two-phase mixture and the indicated level vill show that the pres-surizer is only partially full. Except for closing the PORV block valve, operator actions and system response are the same for these breaks as for similar breaks in the loops.
4.0 SMALl. BREAKS k'I~HOUT AUXILIARY TEEDWATER There are three basic claraes of break response for small breaks without auxiliary feedws'_er.
These are:
l l
1.
Those breaks capable of relieving all decay heat via the break.
}
2.
Breaks that relieve decay heat with both the HP1 injection and j
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.
l i
PAGE DATE:
12-3-80 L.
3W?-20307 (6-7M BABCOCK & WILCOX 28 weitas powlt GtNt4 Aho% DsvtStr.
TECHNICAL DOCUMEN~
"-n:25 1-7
.1
!.OCA's Large. Enough to Deoressuri:e Reactor Coolant Systen Class 1 (Curve 1 of Figure 5), RC sys:e= pressure decreases s cothly throughou: the ::ansient. For the larger breaks in this class, CF-actuation and LP! injection will probably occur. For the smaller breaks of this class only, CFT actuation will occur. Auxiliary feedwater injection is no: necessary for the short-ter: stabili:ation of these breaks.
~he pressuri:er level for :his ::ansient rapidly falls off scale. Operator action and plant response are simila to those described for this class of breaks with a feedvater supply.
4.2 LOCA's Vnich 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 =inutes).
With the HPI's on, a slow systes depressuriza: ion will be established coincident with the decrease in core decay heat.
No Cy actuation is expected. Auxiliary feedwater is no: necessary for the short-te=
stabilization of :hese 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 ESTAS signal, verify completion of all ESFAS functions, and :ry to establish secondary side cooling. Balancing of :he P.?! must also be performed.
If steam generator feedvater cannot be obtained and RCS pressure is increasing, the operator should open the PORV and provide all the HP!
and makeup capability possible. The goal is to depressurize and cool the core with the ECCS, the PORV, and :he break.
If secondary side cooling is again es:ablished, the operator should verify natural circulation, and if unavailable, bu=p a RC pu=p to complete RCS cooldown with :he steam generators. At this point, the PORY can be closed, the sys:e= refilled, and long-ter: 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 5) breaks. Once the SG secondary side inventory is boiled off,
system repressurization vill occur as the break is not capable of removing all the decay heat being generated in the cor Syste:
repressurization to the FORV or the pressurizer Mafety Nes will occur for s= aller breaks in this class. For the ":ero" oreak case, repressurization to the PORY vill occur in the first five minutes.
Operator action is required within the first 20 minutes to ensure core coverage throughout the transient. For the 117-FA lowered-loop plants, this action can be either manual actuation of the auxiliary feeduster system or the RPI system.
PAGE 33 DATE:
12-3-80 J
BWNP-20007 (6-76)
BABCOCK & WILCOX uu m NUCL1aa powte otwtaanom DivistoN 74-11:2501-00 TECHNICAL. DOCUMENT The establish =ent of auxiliary feedwa:er will rapidly depressurize the RCS to the ESTAS actuation pressure, and syste= pressure vill s:abi-lize at either :he secondary side SG pressure or at a pressure where
- he HFI equals the leak rate. Upon receipt of the low pressure ESFAS signal, the operator must trip the RC pu=ps.
For the Class 3 breaks, pressurizer level response will be as shosm in Figure 6.
The =ini=u= refill :ime for the pressurizer is that for the "zero" break and is shown in Figure 6.
Af ter initially drawing inven-tory fro = the pressurizer, the syste= 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 depressurization will resul: and cause an ou: surge fro = the pressurizer. Complete loss of pressurizer level =ay result.
For :he smaller breaks in Class 3 which result in a syste: repressuri-zation following the actuation of the HPI syste=, pressurizer level will increase and then stabilize.
Without auxiliary feedwater, both the hot and cold leg te=peratures will saturate early in the transient and, for the Class 1 and 2 breaks, will re=ain saturated. For the Class 3 breaks, once auxiliary feedwater is established, the cold leg temperatures will rapidly de-crease to approxi=ately the saturation temperature corresponding to the S0 w eendary side pressure and will re=ain there throughout the re=ainder of the trensient. Hot leg te=peratures will re=ain saturated throughout the event.
3e opera:or needs to =anually initia:e all ESFAS actions, balance HPI flow, and attempt to restore secondary side cooling. In the meanti=e, he should actuate the makeup pu=p and open the PORV in order to cool the core and limit the RCS repressurization. Once feedwater is available, he can close the PORV and continue the RCS cooldosm and depressurization with the stea= generators. If natural circulation has not' been established, he can bump a RC pu=p to cause forced circu-lation. The goal is to depressurize to where the ECCS can refill the i
RCS and guarantee long-ter= cooling.
i 4.4 Small Breaks in Pressurizer I
l See the writeup for small breaks in pressurizer with feedwater.
'S=all breaks in the pressu'.izer will differ from those in the loops in the same manner as those previously described in the section address-l ing small breaks in the pressurizer with auxiliary feed.
l 5.0 TRANSIENTS WITH W TIAL RESPONSE SIMILAR TO A SMALL BREAK l
Several transients give initial alarms similar to small breaks. These l
transients will be distinguished by additional alar =s and indications
[
or subsequent system response.
i l
DATE:
12-3-80 2;
PAGE
SWN?-20007 (6-75)
BABCOCK & WILCOX
=gatae powte otNtaatso% Divtom TECHNICAL DOCUMENT 7 -n:250:-00 Overcooling transien:s such as stea: line breaks, increased f eedwa:er flow, and stea: genera:or overfill can cause ROS pressure decreases with low pressure reactor ::1p and ESFAS actuation.
But stea= line breaks actua:e low stea: pressure clar=s for the af fected stea:
generator, and stea= generator overfills result in high stea= gene-ra:or level indications. The overcooling transien:s will repressurize the primary syste: because of HPI actuation, and will return to a subcooled condition during repressurica: ion.
The i==edia:e a:: ions for bo:h overcooling and s=all break transients are the same, in-cluding tripping of the RC pumps.
The operator will recognice overcooling events during repressuri-cation, if not ooner, and is instructed :o thro::le HPI and res:ar:
the RC pumps, if subcooled conditions are established, by the s=all break operating instructions.
A loss-of-feedwater transient will result in a high reac:ct syste pressure alar: 'sut does not give an ESTAS 3ctuation alar =.
A loss of intel, rated control syste: power transient starts with a high RC pressure trip.
After the reactor trip, this beco=es an overcooling :ransient and will give low reactor syste= pressure and possible ESFAS actuation. S:ea generator levels re=ain high and the syste: beco=e.s subcooled during repressurization.
Design features of the B&W NSS ' provide automatic protection during :he early par: cf small break transients, thereby providing adequate ti=e for small breaks to be identified and appropriate action taken to protect the syste. The only pro =pt =anual operator action required is :o trip the RC pu=ps once the low pressure ESFAS signal is reached.
6.0 TRANSIENTS
- HAT MICH* INITIA*E A LOCA There are no anticipated transients that =ight initiate a LOCA since i
the PORV has.been reset to a higher pressure and will not actuate during anticipated transients such as loss of =ain feedwater, turbine trip, or loss of of f site power.
However, if the PORY should lif t and f ail to reset, there are a nu=ber of indications which differentiate this transient fro = the anticipated transients -identified above. These include:
o ESFAS actuation o Quench tank pressure / temperature / level alarms -
o Saturated pri=ary syste= 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 s=all breaks other than a =alf unctioning PORY PAGE 25 DATE:
12-3-80
o
'BL7-20007 (6-76)
BABCOCK & WILCOX
,,v
. t.
NUCitas powta stNtaaboM Division TECHNICAL DOCUMENT n- = 501-Oc occur af ter a transient, they can be identified by initially decreas-ing RCS pressure and convergence to saturation conditions in the reactor coolant. 5=all break repressuriza: ion, if it occurs, will follow saturation conditions. By re=aining aware of whe ther the reae:or coolant re=ains subcooled or becomes sa:urated af:er tran-sients, the operator is able to recogni:e when a s=all break has occurred.
7.0 HPI *HRo m ING Maintaining adequate core cooling is the most important concern during a small break LOCA. This requires keeping the core covered with subcooled reactor coolant. Becruse of ins:rument errors, it is necessary to maintain a 500F subcooling margin to sa:uration. During forced flow (ROPs on) the amount of subcooling can be determined by the hot and cold leg RTDs. However, with 3 forced flow, the core exit thermocouples must be used to deter =ine the amount of subcooling.
During a small break LOCA, suf ficient HPI flow is necessary to assure adequate core cooling. Theref ore it is critical that H?! flow is g throttled unless the reactor coolant is greater than 50oF subcooled.
Beyond 500F subcooled, certain cases require throttling of HPI flow to tvoid exceeding other limits. *hese cases are A. to ensure reactor ressel integrity, B. prevent pressurizer level fro: going off-scale high, and C. allow termination of HPI flow once LPI cooling is assured. Each of these cases is discussed in detail as follows:
A.
Reactor Vessel Integrity The RCS pressure / temperature combination must be kept wi:hin cer-tain li=its to assure reactor vessel inte; '..y.
These limits are dependent on whether there is 1. Forced Flow, or 2 3 Forced flos:
1.
Forced Flow l
As long as the reactor coolant pumps (ROPs) are running, the RCS pressure and temperature must be kept within the nor=al technical specification NDT limits (Region I & II of Figure 1).
With RCPs running, any cold Leg RTD can be used to determine the temperature for comparison to the NDT limit.
Also, while the initial temperature drop will be deter =ined by the size of the break, as soon as temperature control is achieved the rate of cooldown should be limited to < 100"F/hr.
2.
No Forced Flow if the RCPs are NOT running, the RC pressure / temperature cos-bination must be kept within the No Forced Flow region of Figure 1, (Region II). The " Interim Brittle Fracture Limit" of Figure 1 is based on an analysis of ther=al shock to the PAGE 26 DATE:
12-3-80
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.... :. v u o BABCOCK & WILCOX scitas *o* te otNf ta?'o% omsic TECHNICAL DOCUMENT 7"-112:5:1-3:
reactor vessel wall. Wi:h n RCPs running, cold "?1 wa:e: could enter the down:o=er and s:rea: down the rea::or vessel wall vi:h relatively little =ixing with the ven: valve flow.
The resulting ther=al shock :o the rea::or vessel vali could exceed the bri::le fra::ure 11:i: if the RCS pressure is no: reduced below an allow-able value within a certain a=ount of :i=e.
Recen: analyses in-dicate tha: :hro:: ling the EPl flow to =aintain :he "Interi:
Brittle Fra::ure L1=i:" of Figure i vill =ee: :he required pressure 11=its.
The "In:eri: Bri:tle Fra::ure Limit" of Figure 1 is based on several conservative assu=ptions, consequently, s=all viola: ions of this li=i: are = ore :olerable than similar viola: ions of :he 500 subcooling =argin.
Eovever, if the "interi: Brittle Fra::ure Li=1:" is exceeded, the ROS pressure should be reduced to regain the no forced flow operating region as quickly as possible.
Several a:: ions are i=por: ant to ensure tha: the brittle fracture 11=1: is not exceeded during a r=all break LOCA. These include a.)
con:inuous =oni:oring of ther=ocouple te=peratures, b.) =aintaining RCS pressure cont:cl and, c.) restoring natural circula: ion if pos-sible.
a) Monitorine Ther=occuole Te=ceratures With no ROPs running, the average of the five highes: the r:c-couple te=perature readings should be used to et ;er=ine the RO te=perature for Figure 1.
This will assure tha: :he 50*r sub-cooling =argin is =aintained and tha: the brittle fra::ure li=i:
is not exceeded.
b) RCS Pressure Control With no RCPs running, thro:: ling the HP1 flow is the bes:
=ethod for gradually reduci_ng RCS pressure. Also, withou:
pri=ary to secondary hea: ::ansfer, the rate of cooldown is dependent on RP1 cooling :hrough :he break (opening :he PCRV is ne:essary only if the break is so s=all that RCS pressure begins increasing). The efore, careful and consistent thro::-
ling of HP1 flew is the only available =eans to ensure tha: the RC pressure /te:perature co=bina: ion re =ains within the no for:-
ed flow eperating region of Figure 1.
t c) Restoring Natural Circulation If RCPs are tripped, natural circulation should be obtained :o
~
provide so=e HP1 =ixing as well as providing good hea: transfer from the primary to secondary coolant. With natural circula-tion, the cold HPI water vill mix with cold leg flow and reduce the ther=al shock to the reactor vessel. However, the RC pres sure/ te=perature should still be =aintained within the no forced flow opera:ing region of Figure 1.
Natural circulation DATE:
FAGE 12-3-80 27
BZNP-20007 (6-76)
BABCOCK & WILCOX myttgaa towtt GENetAf tok Division 74-11:2501-0C TECHNICAL DOCUMENT is verified by (1) cold leg temperature is saturation temperature of secondary side pressure and (0) primary iT (Tgo: -TCold) becomes constant at 500F. The brittle fra:ture concern is eliminated entirely when RCPs are running and ROS P/T is main-tained within Tech Spec NDT limits. Therefore, as soon as 500F subcooling is obtained in the RCS, a ROP should be restarted.
Then the reactor vessel downcomer pressure /te=perature should be kept within the normal NDT limits.
3.
Prevent Pressurizer Level From Goinc Off-Scale High Provided that the RCS is at least 500F subcooled, RPI flow should be throttled if necessary to prevent the indicated pressurizer level from going of f-scale high. Under these conditions, the primary system is solid. Continued HPI flow at full capacity may result in a solid pressurizer and would result in a lifting of the PORV and/or the pressurizer code safety valves. This may in turn lead to a LOCA. Thus, HPI flow should be throttled to maintain a stable inventory in the RCS. However, if the 50oF subcooling cannot be maintained, the HPI shall be immediately reactivated.
C.
Termination of HPI Flow Once LPI Cooling is Assured 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, termination of the HPI is allowed.
However, this action should only be taken if the flow rate through each line is at least 1000 gpm* and the situation. has been stable for 20 minutes. The 20-minute time delay is included to ensure
[
that the system will not repressurize and result in a loss of the l
LPI fluid. In the event of a core flooding line break, the LPI fluid entering the broken core flooding line will not reach the ve s sel,. Thus, in order to ensure that fluid is continually being l
l
. 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 two HPI pumps and ensure that upon termination of the HPI pumps, adequate flow is being delivered to the RV.
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- For Arkansas Nuclear One use 2630 GPM to either injection line.
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.DATE:
^6 12-3-80 28 l
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TECHNICAL DOCUMENT 7 -n :2501-x FIGURE 2 PRESSURE VS TDIE-SMALL BREAKS WITH AUXILIARY FEEDWAT2R 2500 -
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12-3-80 PAGE 30
sar-:oo:: ce-:e SABCOCK & WILCOX
" # 8E8 wantal POwet GtNtaafica D'vi$ic%
TECHNICAL DOCUMENT N-n:2E-x FIGURE 3 PRESSURIZER LEVEL VS TIV.E-S\\ TALL BREAK 5 WITH AUXILIARY FEEDWATER 100 p-q I
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1213-80 PAGE 11
5.:N?-2000- (6-76
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TECHNICAL DOCUMENT FIGURE 4 PRE 55URIZER LEVEL V5 TDfE FCR 55 TALL SREAK IN PRE 55UR::IR l
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12-3-80 PAGE 32 I
S'a5?-20007 (e-76 BABCOCK & WILCOX Numsta NWO.!As Powie GENetatiCN D'v!5 ton 7"-
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TECHNICAL. DOCUMENT FIGURE 3 SYS 3 1 PRESSURE VS TIME-S>RLL BREAKS K/O AUXILIARY FEEDWATER 2500 "ZER0" l LEAK 3
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DATE:
12-3-80 PAGE
BRC - 20007 (e-76, BABCOCK & WILCOX "U" 8 I' huClas Powte GINgaailON Dtvinich 7 -2 22 m - C TECHNICAL DOCUMENT FIC'JRE 6 PRESSURIZER LEVEL VS TIFE-CLASS 3 l
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DATE:
12-3-80 PAGE 34
BWNP-20007 (e-76)
B ABCOCK & %VILCO.X hu:Lt Aa Powts Gtwitatio% orv@ow TECHNICAL DOCUMENT 7 -112:501-00 APPENDIX A INADEOUATE CORE COOLING - DESCRIPTION OF PLAN
- BEHAVIOR 1.0 IN*RODUC* ION Following a loss-of-coolan: accident (LOCA) in which the reactor trips, it is necessary to remove the decay heat from :he reac cr core to prevent damage. Core heat removal is accomplished by supplying cool-ing water to the core. The water which is available for core cooling is a portion of the initial reactor coolant system (RCS) water inven-tory plus any water injected by the emergency core cooling syste=
(ECCS). The heat added :o 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 becomes uncovered, cool-ing for those regions will be by forced convection to superheated steam which is a low heat transfer regime. Continued opera: ion in the stea=
cooling mode will resul: in elevated core te=peratures and subsequent core damage.
2.0 LOSS OF RCS INVENTORY WI""E REACTOR COOI. ANT PL PS OPER,. TING Vith the RC pumps opera:ing 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 e high void fraction for cer:ain small breaks with the RC pumps on, a RC pump trip by any means (i.e., loss of of f site power, equipment failure, etc.) at a high void fraction during the small break ::ansient may lead - to inadequate core cooling. That is, if the RC pumps trip at a time period when the syste= void fraction is greater than approximately 70%, a core heatup will occur because the amount of water lef t in the RCS would not be sufficient to keep the core covered. The cladding temperature would increase until core cool-ing 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 pumps upon receipt of a low pressure
' ESTAS signal is required in order to ensure that adequate core cooling is provided. Following the RC pump trip. the small break transient concerns about inadequate core cooling will be the same as described in the previous section.
DATE:
PAGE 35 12-3-80
BWNP-20007 (6-76>
BABCOCK & WILCOX
.t.
NUCitas powts otNteAfsom Divi $ tom TECHNICAL DOCUMENT 7'-11:2531-00 If the RC pu=ps can no: be tripped by the operator, :he :on:inued forced circulation of fluid throughout the RCS will keep the core cooled. Bovever, if li::le or no ECCS is being provided to the RCS, the fluid in the RCS will eventually become pure stea= due to the con:inued energy addition to the fluid provided by the core decay heat.
Under these circumstances, an inadequate core cooling situation will exist. Since :he heat re= oval process under forced circulation is better than the stea= cooling mode described below for the pu=ps off situation, the operator ac: ions and indications described in :he subsequent section are sufficient for inadequate core cooling wi:h the RC pu=ps operating.
3.0 LOSS OF RCS INVENTORT WITHOL~ REACTOR C001. ANT PUMDS OPERATINC Without the RC pu=ps operating, the cooling of the core is acce=plished by keeping the core covered with a stea=-water =ixture.
As the fluid in the core is heated, some of it or all of it =ay be turned to stea=.
If insuf ficient cooling water is available to =aintain :he stea=-water mixture covering :he core, the core exit fluid te=peratures will begin to deviate fro = the saturation te=perature corresponding :o the pressure of the RCS.
One i==ediate indication that inadequate core cooling =ay exist in the core is that the te=perature of the core exit ther=ocouples and hot leg RTD's are superheated. A: 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 bu: is not, by itself, a cause for ex:re=e action. If the ECCS syste=s are f unctioning nor= ally, the te=peratures should return to saturation without any actions beyond those outlined for a s=all break. For incere ther=ocouple te=perature indicating superheated conditions, the operator _ should (a) verify a=ergency cooling water is being injected through all HPI nozzles into the RCS, (b) initiate any additional sources o' cooling water available such as the standby =akeup pu=p, (c) verify the s:ea= generator level is being =aintained at the e=ergency level (d) if stea= generator level is no: at 95% of opera:ing range raise level to the 95: level, (e) if the desired stea= generator level cannot be achieved, actuate any additional available sources of feed-water such as startup auxiliary feedwater pu=p, (f) establish 1000F/ hr cooldown of RCS via stea= generator pressure control, (g) open core flooding line isolation valves if previously isolated, and (h) if RC pressure increases to 2300 psig open the pressurizer PORY to reduce RC pressure and reclose PORV when RC pressure falls to 100 psi above the secondary pressure. These actions are directed toward depressurization of the RCS to a pressure at which the ECCS vater input exceeds core stea= generation. The align =ent of other sources of cooling water is the recognition that the injection of the HPI syste= alone is not sufficien: to exceed core boil off.
DATE:
12-3-80 pAGE 36
Sk'NF-20007 (6-76; BABCOCK & NVILCCD(
waea no-en on~eemo~ omsio~
TECHNICAL DOCUMENT 7;-11:: sol-Oc lf the incere thermocouple indications reach Curve #1 on Figure 3 in Part 1, the peak fuel cladding temperature has reached approxima:ely 14000F. Above this temperature level there is a potential for cladding rupture. Also, :he zircaloy cladding water reaction will begin to add a significant amount of hea to the fuel cladding thereby greatly increasing the possibility of core structural damage unless adequate core cooling is restored. Non-condensible gas for=ation will increase rapidly from this level of fuel clad :esperature.
For incore thermocouple te=pera:ure indications at or exceeding Curve
- 1 on Figure 3 in Part 1, the operator should (a) start one RC pu=p in each loop, (b) depressurize the s:eam generator as rapidly as possible to 400 psig or as far as necessary to achieve a 1000F decrease in satu-ration temperature, (c) immediately continue the plant cooldown by main-taining a 1000F/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 ne RC pu=ps will provide forced flow core cooling and will reduce the fuel cladding temperatures. The rapid depressurization of the steam generators will help to depressurize the primary system to the point where the core flooding tanks will actuate. Stopping the depressurization at 400 psig (or at a reduction in saturation temperature of 100 F) will main:ain the tube to shell te=pera:ure dif ference within the 100 F design limit.
The continued cooldown to 150 psig will reduce the prt=ary sys:en pres-sure to the point where the Low Pressure Injection System can supply cooling. The opening of the PORV will also help to depressurize the primary system. The PORV should be closed when the primary pressur9 is within 50 psi of the secondary pressure and then should only be uped as necessary to maintain the primary system pressure at no greater enan 50 psi above the secondary system pressure. This method of operation vill minimize' the loss of water from the primary system through the PORV.
If the incore thermocouple readings reach Curve #2 on Figure 3 in Par:
1, the peak cladding temperature is approximately a: the 18000F level.
l This is a very serious condition. At this level of clad te=perature, significant amounts of non-condensible gas are being generated and core damage may be unavoidable. Extreme measures are required by the opera-cor to prevent major core damage. The goal of these actions is to i
depressurize the RCS to a level where the core flooding tanks will l
fully discharge and the LPI system can be actuated thus providing prompt core recovery. The operator should (a) depressurize the steam l
generators as rapidly as possible, (b) start the remaining RC pu=ps and (c) open the PORV and leave it open.
l l
l i
DATE:
12 3-80 pAGE 37 L
a 5*.7.~7 -1000 ~ ( t - 7 6 ;
BABC a..O,o K. L.~W.ILCOX C
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TECHNICAL. DOCUMENT n-n r 531-x For a very s=all or non-LOOA even:, the core decay hea: re= oval is acceeplished via the stea: generators. If tha: hea: re= oval is decreased or lost, the natural circulation of fluid withis :he ROS =ay be reduced or stopped. The loss of natural circulation for core cool-ing vill eventually boil off the remaining water inven:ory in the core and lead to inadequa:e core cooling and eleva:ed core te=pera:ure.
Indica: ions of loss of stea: generator heat sink include (a) a icw level in the steam genera:or with low stea= pressure, (b) te=pera:ure indicators in ho: legs show satura:ed ta=pera:ures, (c) increasing ROS pressure. The operator should ::y to establish emergency feedva:e: as quickly as possible and i==ediately actuate the EP1 syste= to restore natural circulation and RCS hea: removal. If auxiliary feedwa:e is no: available and there is no break in the ROS, the system will re-pressurize and de:ay heat will be removed by opening the PORV and maximizing EPI addition.
5 I
DATE:
12-3-80:
PAGE 33
. -.