ML19312C798: Difference between revisions
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r TABLE 2: RC PLMP RESTART CRITERIA I 2,3 CONDITION FOR lAllCH TYPICAL EVENTS FOR INSTRUCTION LOCATION DISCUS $10N A PIN RESTART IS ALLOldED lAllCH A RCP RESTART IN SMALL BREAK GUIDELINES IS ALLOWED (SECTION) | r TABLE 2: RC PLMP RESTART CRITERIA I 2,3 CONDITION FOR lAllCH TYPICAL EVENTS FOR INSTRUCTION LOCATION DISCUS $10N A PIN RESTART IS ALLOldED lAllCH A RCP RESTART IN SMALL BREAK GUIDELINES IS ALLOWED (SECTION) | ||
Regain Coolant Subcooling 1. Small Leak 4.3.4.3.2 Following any reactor trip event during | Regain Coolant Subcooling 1. Small Leak 4.3.4.3.2 Following any reactor trip event during | ||
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G: HOT LEG 'B' LOOP-RC PUMP TRIP 6: HOT LEG (PRZR LOOP) - NO TRIP 0: HOT LEG 'B' LOOP-N0 TRIP | G: HOT LEG 'B' LOOP-RC PUMP TRIP 6: HOT LEG (PRZR LOOP) - NO TRIP 0: HOT LEG 'B' LOOP-N0 TRIP | ||
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5 1500 - | 5 1500 - | ||
s a e o | s a e o KEY E | ||
KEY E | |||
= | = | ||
O: RC PUMPS TRIP o . | O: RC PUMPS TRIP o . | ||
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= m LaJ A A | = m LaJ A A | ||
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* Z D C = CE" w | * Z D C = CE" w | ||
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~ ~ ~ 1.2 ANS, PUMP OFF -l 0.05 FT2 , 2 HPI'S | ~ ~ ~ 1.2 ANS, PUMP OFF -l 0.05 FT2 , 2 HPI'S | ||
.0 ANS, PUMP OFF 3m. 2000 - | .0 ANS, PUMP OFF 3m. 2000 - | ||
a. | a. | ||
1500 r a- I L | 1500 r a- I L | ||
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1 Figure 3.2 1 | 1 Figure 3.2 1 | ||
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\g - 1500 g- | \g - 1500 g-h 4 5> l s I | ||
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= 5 | = 5 | ||
%.J.,.%s, - 1000 ; | %.J.,.%s, - 1000 ; | ||
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/ | / | ||
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two of four pumps running '(See Figures 3.2,3.3). In both cases, the system is highly subceoled, from a minimum of 30*F to 120*F ' | two of four pumps running '(See Figures 3.2,3.3). In both cases, the system is highly subceoled, from a minimum of 30*F to 120*F ' | ||
) | ) | ||
and increasing at the end of 14 minutes (refer to Figure 3.4). i i | and increasing at the end of 14 minutes (refer to Figure 3.4). i i | ||
~ | ~ | ||
| Line 1,372: | Line 1,355: | ||
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: 2. Method of Analvsis - | : 2. Method of Analvsis - | ||
The analysis method used for this evaluation is basically that de-scribed in section 5 of RAW-10104. Rev. 3, "B&W's ECCS Evaluation Model"1 and the letter J.H. Taylor (B&W) to S.A. Varga (NRC), dated | The analysis method used for this evaluation is basically that de-scribed in section 5 of RAW-10104. Rev. 3, "B&W's ECCS Evaluation Model"1 and the letter J.H. Taylor (B&W) to S.A. Varga (NRC), dated | ||
* July 18, 19782 , which is applicable to the 177-FA lowered-loop plants for power livels up to 2772 MWt. The analysis uses the CRAFI23 code to deve'op the history of the RCS hydrodynamics. | |||
July 18, 19782 , which is applicable to the 177-FA lowered-loop plants for power livels up to 2772 MWt. The analysis uses the CRAFI23 code to deve'op the history of the RCS hydrodynamics. | |||
1 | 1 | ||
+ | + | ||
| Line 1,396: | Line 1,378: | ||
_, , - . ~ . , . _ . | _, , - . ~ . , . _ . | ||
Following tripping of the RC pumps and the subsequent loss-of-forced circulation, the system will collapse and separate. | Following tripping of the RC pumps and the subsequent loss-of-forced circulation, the system will collapse and separate. | ||
The residual liquid will then collect in the reactor vessel and the loop seal in the cold leg suction piping? For this period of the transient, the Wilson bubble rise model is utilized. | The residual liquid will then collect in the reactor vessel and the loop seal in the cold leg suction piping? For this period of the transient, the Wilson bubble rise model is utilized. | ||
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. 1 | . 1 | ||
local power rate (kw/ft) analyzed is not expected to occur during normal plant operation. Furthermore, use of an adiabatic heatup assumption neglects any credit for the steam cooling that will occur | local power rate (kw/ft) analyzed is not expected to occur during normal plant operation. Furthermore, use of an adiabatic heatup assumption neglects any credit for the steam cooling that will occur during the core refill phase and also neglects the effect of any radiation heat transfer. Using a decay heat power level based on 1.2 ANS at 1500 seconds, the cladding will heatup at a rate will be 6.5 F/S under the adiabatic assumption. With a core uncovery period of j l | ||
during the core refill phase and also neglects the effect of any radiation heat transfer. Using a decay heat power level based on 1.2 ANS at 1500 seconds, the cladding will heatup at a rate will be 6.5 F/S under the adiabatic assumption. With a core uncovery period of j l | |||
600 seconds and the adiabatic heatup assumption, cladding temperatures will exceed the criteria of 10CFR50.46. Use of a more realistic heat - | 600 seconds and the adiabatic heatup assumption, cladding temperatures will exceed the criteria of 10CFR50.46. Use of a more realistic heat - | ||
i transfer approach with the extreme power shape utilized for this eval-untion is also expected to result in cladding temperature in excess of the criteria. In order to ensure compliance of the 177 FA lowered loop plants to the criteria of 10CFR50.46 a prompt tripping of the RC pumps is required. Section B. demonstrates that a prompt trip of . | i transfer approach with the extreme power shape utilized for this eval-untion is also expected to result in cladding temperature in excess of the criteria. In order to ensure compliance of the 177 FA lowered loop plants to the criteria of 10CFR50.46 a prompt tripping of the RC pumps is required. Section B. demonstrates that a prompt trip of . | ||
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: 7) The above conclusions are applicable to both the B&W 177 FA lowered and raised loop NSS designs. | : 7) The above conclusions are applicable to both the B&W 177 FA lowered and raised loop NSS designs. | ||
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8 . | |||
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characteristics. Plant recovery would be more difficult, dependence. | characteristics. Plant recovery would be more difficult, dependence. | ||
on natural circulation mode while achieving cold shutdown would be | on natural circulation mode while achieving cold shutdown would be highlighted, manual fill of the steam generators would be required, and so on. However, all of these drawbacks can be accouanodated since | ||
highlighted, manual fill of the steam generators would be required, and so on. However, all of these drawbacks can be accouanodated since | |||
( none of them will on its own lead to unacceptable consequences. Also, restart of the pumps is not precluded for plant control and cooldown | ( none of them will on its own lead to unacceptable consequences. Also, restart of the pumps is not precluded for plant control and cooldown | ||
'once controlled operator action is assumed. Out of this search, | 'once controlled operator action is assumed. Out of this search, | ||
Latest revision as of 18:49, 21 February 2020
| ML19312C798 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/30/1979 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| IEB-79-05C, IEB-79-5C, NUDOCS 7912191022 | |
| Download: ML19312C798 (50) | |
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O O ATTACIDfENT 3 _ GUIDELINES FOR OPERATOR ACTION O e
j GUIDELINES FOR OPERATOR ACTION I. Introduction Guidance for operator action, during both LOCA and non-LOCA events, to account for the impact of the RC pump trip requirement of IE bulletin No. 79-05C, have been developed and are presented below. The general intent of these additional instructions is as follows: ,
- 1. To establish the basis and criteria for a RC pump trip and _i
- 2. To identify plant conditions for which a restart of the RC Pumps, if
- . tripped, is permissable.
I Section VI provides the " Operating Guidelines for Small Breaks" updated to include the impact of the RC pump trip requirements. These guidelines, in general, apply to any abnormal event where a RCP trip is required and will be used as the basis for revisions to emergency operating procedures and operator training. II. Basis and Criteria for a RC Pump (RCP) Trip B&W analyses of small loss-of-coolant accidents, v ch the RC pumps operative, indicated that the primary reactor coolant conditions evolve to high void fractions during the initial stages of the transient when the system pressure is still relatively high. The consequences of these postulated events with continouous RC pump operation are acceptable as effective core cooling is maintained due to the forced circulation of reactor coolant. For a certain
, range of small breaks, however, a RCP trip (by any means such as loss of power or operator action) at a time when the coolant void fraction is excessively high can lead to core uncovery and a potential for cladding temperaturas in excess of 2200F.
To preclude the potential consequenc,e of an untimely RCP trip, the RCP's will be promptly shutdown when R'CS conditions indicate a small break in
. this size range may be in progress. This action ensures safe plant con-ditions as demonstrated by past small break analyses, under Appendix K assumptions, wherein the RC pumps were assumed inoperative early during the transient.
In the interim, until design changes can be made to automate the RCP trip, operating procedure will require that the operator trip the RCP's immediately following ESFAS actuation due to low RC pressure (< 1600 psig). Tabic 1 outlines the general diagnostic and confirmatory actions which will be required in addition to other immediate actions in present procedures. These imediate actions apply to any abnonnal event which results in automatic ESFAS actuation on low RC pressure and will be memorized by reactor operating personnel during training programs. As indicated above, a prompt trip of the RC pumps is required in order to maintain demonstrated conformance to 10CFR50.46. To provide good assurance that the operator will trip the RC pumps when required, the ' pump trip criteria (low pressure ESFAS actuation) was chosen over other possible candidates because it is a clear, simple, and early indication that a small LOCA may be in progress. The visual indication and alarms in the control room following ESFAS actuation also alert the operator to the status of the plant, and no decision process or continuoas monitoring by the operator is required to decide that an RC pump trip is necessary. With procedure changes consistent with Table 1 and additional training, failure of the operate to, initiate an RC pump
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trip when required is believed to be remote.
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. l Table 1: IMMEDIATE ACTIONS REQUIRED FOLLOWING ESFAS ACTUATION i l
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- 1. Criteria for RCP Trip '
, Upon automatic actuation of the ESFAS due to low reactor coolant system ;
pressure. RC pump operation shall be promptly terminated.
- 2. Immediate Action A. Upon receipt of an ESFAS actuation (indicated via audiable and visual alarms within the control room) the operator shall immediately ver fy -
that RC pressure is less than the low pressure ESFAS setpoint via ! examination of wide range RC pressure instrumentation or ESFAS Trip Status Indication, if available. B. If RC pressure is less than the low pressure ESFAS setpoint, RC pump
. operation shall be immediately terminated by manual depressing the in-dividual RC pump trip switches in the control room.
NOTE: If the ESFAS has been actuated due to high RB pressure, the operator shall monitor RC pressure and trip the RC pumps if pressure decreases below the ESFAS setpoint. C. The operator shall immediately verify that the RC pumps are tripped by visualexaminationofRCpumpstatusindications(statuslights, motor , current,etc.). D. Following a trip of the RC pumps, the operator shall verify that the
. auxiliary feedwater system has been acteated and that SG level is controlled to the emergency high level control setpoint. to ensure establishment of natural circulation.
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III. Criteria for RCP Restart Plant control following abnormal events, including small breaks, is greatly improved if the RC pumps are operative. With forced circu-lation of reactor coolant, the steam generators and associated auxiliary systems are more effective in removing the primary system stored energy , and decay heat. The plant is also placed in a more " normal" mode of l operation where more familiar pressure / temperature control procedures can be employed by operating personnel. Therefore, to compliment the RC pump trip criteria provided in Section II, conditions under which an RC pump restart is allowed have also been identified. These conditions _l cover both LOCA and non-LOCA events and have been carefully chosen to j preclude the development of excessive void fractions for small breaks l where an RC pump restart is allowed. . l Table 2 lists the conditions under which a RC pump restart is allowed. l For each condition, typical events for which they apply and a brief l discussion of the basis for the RC pump restart issprovided. It should be l noted that a RC pump restart is not allowed unless feedwater is available : l to at least one steam generator. A cross-reference to the appropriate l sections of the small break guidelines where specific infonnation can be { found is also given. Furthermore, the criteria given in Table 2 are i <
; not new as each was previously issued in past small break guideline submittals. l B&W has reviewed the guidelines in light of the break size and system conditions ;
for which a RC pump trip is required and has confirmed that the RC pump restart guidance is still appropriate. As indicated in Table 2, system repressurization and the establishment of subcooled conditions are specified for use on non-LOCA events as criteria for which a RC pump restart is allowed. For these abnormal events, restart of the RC pumps is recommended by B&W when the Pump Restart criteria is satis-fied to aid in plant recovery and control. Emergency procedures for non-
LOCA events, for which a RC pump trip may be initiated. Will thus be revised to include the pump restart criteria. 9 e amm s** e 9 e O d 99 ',M q
r TABLE 2: RC PLMP RESTART CRITERIA I 2,3 CONDITION FOR lAllCH TYPICAL EVENTS FOR INSTRUCTION LOCATION DISCUS $10N A PIN RESTART IS ALLOldED lAllCH A RCP RESTART IN SMALL BREAK GUIDELINES IS ALLOWED (SECTION) Regain Coolant Subcooling 1. Small Leak 4.3.4.3.2 Following any reactor trip event during
- 2. Small Break within which the RC pumps become inoperative
- 1. P-T conditions indicate capacity of HPI sys. (loss of power due to siatural causes/
coolant is 2,50F subcooled. 3. Isolated Small Break equipment failures or due to a deltterate
- 4. Non-LOCA Overcooling / trip initiated by the operator), the RC depressurizing event pumps can be restarted if RC conditions *
, 5. Loss-of-Offsite Power -
are stablitzed and at least 50F of sub-Event cooling is indicated for the entsting P-T state. If subcooled conditions are indicated, the primary and secondary sys. ' tems are directly coupled (ie, decay *
- heat remosal via natural circulation);
and if a bee.ach of the primary pressure boundary is present also, the resulting leak will to within the capacity of the ECCS systems. The operator should restart the RC pumps (1 in, each loop) return ,
. to low SG Level control, and proceed with a plant cooldown or maintain the plant at hot shutdown if the initiating event is correctable and a return to power i operation possible.
NOTE: The subcooling criteria will be , the principle indicator for a g RCP restart for non-LOCA events. j i i Repressurization
- 1. Stable or increasing 1. Small Break within capacity 4.3.4.4.1 Certain ssall breaks viil result in a pressure with PRCS > of HPIS system re,;ressurization due to unmentary 1
- 1600 psig. 2. Overcooling /Depressurtration loss of the SG as a condensor for primary event systes, steam (ie, the HPIS is' refilling
- 3. Isolated Small Break- the system and a steam bubble is trapped ,
within the hot legs above the SG tubes condensing surface). Small breaks which produce this primary system behavior are sufficiently small such that high void fractions will not evolve if the RC pumps are restarted. A RCP restart , is thus allowed; this action will equal-1re primary and secondary pressures and temper'.tures and couple the primary and secondary systems such that an orderly cooldown and depressurization of the RCS canf.o accomplished. Section 4.3.4.4.1 of ti.:. small break guidelines would
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TABLE 2 CONT'D 2 C0hD TION FO' 1AllCH .3 TYPICAL EVENTS FOR INSTRUCTION LOCATION DISCUS $10N A Ptw RE/.Wf IS ALLOWED net!CH A RCP RESTART IN SMALL BREAK GUIDELINES 15 ALLOWED (SECTION) , a> ply to a very small break where a sys-tem repressurization would occur early (ie, prior to initiation of the second-ary system depressurtration). A RCP restart and resulting drop in the primary system pressure to that of the second- ,
, ary side may allow the HPIS to estabitsh a subcooled primary system. System repressurization above the low pressure ESIAS setpoint for non-LOCA events is also an acceptable condition for an RC pump restart. In most cases. Increasino RC pressure will also tend to re-establish the reactor coolant subcooled margin '
which as indicated above, is the principle indicator for a RCP restart for non-LUCA event. A pump restart. when system pressure is aoove the ESFAS setpoint when the 50F subcooled margin is not yet - estabitshed, is permissable since small '
. breaks for which a RC Pump trip is re-quired will not produce the system behavior. ,
- 2. Increasing system pressure Small Break 4.3.4.4.2 4.3.4.4.2 of the small break guidelines wherePRCS>+600(psig) applies during the cooldown process where during coolilown process. the secondary pressure has been manually .
reduced below normal control (hot shut-s down) setpoints. A pump bump procedure is stipulated. The intent of this action . is to mix the system so that steam can ' be condensed to allow a system refill. If a refill and subcooled conditions are not established. the 600 pst decrease . In primary system pressure *!!1l prevent high RCS vold fractions with an RCP restart per the guidance provided. Final Transition to LPI Cositne 4 Stablired pressure with Sma11 Break 4.3.4.4.3 For certain small breaks. a primary system PSS< 100 psig and PRCS refill may not be possible until low
> 250 psig primary evstem pressuses are achieved.
i* Complete wepressurtaation may be impeded due to steam trapped within the upper hot leg piping. A bump of an RCP will depressurize the RCS such that a transition to LPI cooling per Appendix A of the small break guidelines is possible i r L.a Luam' _
TABLE 2 CONT'D CONDITION FOR nGtICH 2.3 . TYPICAL EVENTS FOR INSTRUCTION LOCATION DISCUSSION A PUW RESTART 15 ALLOWED WHICH A RCP RESTART IN SMALL BREAK GUIDELINES IS ALLOWED (SECTION) Continued operation of an RCP is also allowed since the LPI system will elial-nate the potential for further increase in the system vold fraction. Inadequate Core Cooling Small Break N/A Current considerations of the indicatto.:s of and mitigating actions for inadequ-ate core cooling may result in the potential use of the RC pumps under certain condi-tions. Criteria for use of the RC pumps. 1f requi.3d wlIl be developed consistert with the sstedule requirement of item 5 (short tern) of 79-05C. NOTE: 1. An RC Pump recte t is allowed only if feeduster
- is available to at least one steam generator.
- 2. Standard precautions to be observed prior to pump restart.
A. CCW has been maintained or will be reinstated prir- '; starting the RC pumps.
- 8. Seal injection flow has been maintained to all RC pumps.
C. Seal return is maintained or is reinstated prior to starting . pumps. D. Prcs r 250 psig. ~
- 3. Emergency operating Ilmits for continued pump operation.
A.Shaftrunout(vibration)shallnotexceed30 alls. B. Frame vibration as measured on the lower motor mounting flange shall not exceed 5 alls. hh J lben a Ah O O N; K!J m i
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. IV. Operating Guidelines for Small Break Part .I and Part II of the " Operating Guidelines for Small Breaks" have been revised to include the RC pump trip requirement of II Bulletin 79-05C and are attached. This infonnai.icn will serve as the basis for revisions to emergency procedures and additional operator training.
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l V. Guidelines for Non-LOCA Events l Because of the broad spectrum of system conditions covered by the small break guidelines, the operator actions and precautions identified to bring
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the plant to a long term cooling mode apply, in general, to any abnormal event which results in a decrease in RCS pressure. The small break guidelines will thus be utilized to update the emergency procedures for non-LOCA events; at a minimum, the following pertinent sections of the small break guidelines will be incorporated:
- l. RC Pump Trip Criteria and SG Level Control actions to promote natural circulation.
- 2. RC pump Restart Criteria
- 3. HPI Control Criteria
- 4. The need to monitor system subcooling limits.
The items will be supplemented by the additional instructions / precautions to the effect that: ,
- 1. For non-LOCA events, a restart of the RC pumps (1 per loop) and termination
.of SG fill is prudent to minimize system overcooling due to addition of cold AFW to the OTSG's.
Note: The establishment of a subcooled condition (>50F) is a clean indication that a non-LOCA event or a LOCA for which a RCP trip is not required is not in progress.
- 2. HPI should be throttled, when 50F subcooling is established, to avoid a -
pressurizer overfill.
- 3. During severe overcooling events, sufficient HPI water may be added, prior
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to achieving a subcooled condition (> 50F) and a pressurizer level (on-scale), such that the system may evolve to water solid state when the RC temperature recovers to a hot shutdown condition (s 530F).
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Operator action to control primary temperature (via secondary steam pressure control using the turbine bypass valves and/or atmospheric
, dumps) may be required to mr'ntain pressurizer level on scale.
NOTE: The Operating Guidelines For Small Breaks have been modified to include Item 3 above. With operator training in the post-LOCA recovery methods in conjunction with modification of existing emergency procedures based on the small break guidelines, plant recovery and control can be achieved for any _ abnormal event for which an KOP trip is required.
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i l ? \ \ l 1 . ATIACEMENT 4 _ AUTOMATIC RCP TRIP SCHEMATIC 1 G 1 1 i 1
, . s VITAL SOURCE A VITAL SOURCE A CH. I BYPASS 7 )
V > TRIP I RCPAl HPI CH. I TRIP > RCP Al LOW POWER > AND > TRIP i RCPBI RCP Of LOW POWER >
> TRIP 2 RCPA 2
> TRIP 2 RCPS2 f
e
~
VITAL SOURCE B VITAL. SOURCE B JL CH. 2 BYPASS ( ) HPI CH 2 TRIP > ' CCP At LOW POWER > AND -
> TRIP 1 TA STARTdP FEEDER
> TRIP l TS NORM. FEEDER'
> TRIP 1 T8 STARTUP FEEDER 4 i Yu,uUbuemuudLl,f(gh{h0'u ._ ~
ATTACHMENT 4
C00LA':T TE"?Eh?.TU".Es VERSUS TEl.': Sit!;T TI:I (In2 FP, BEGil:ll!'G UF LIFF, 12.2 FT 2000P! E E!l9 P.UPTU7E,, STE!."'.!!!E OREl.K (U::"lTICf.TED)
!!9 PC P""' TP!P) 700 , y i
. P"ESSURIZER E*/PTIES
\
,s HOT LEG BEGINS TWO PHASE
\ U 600 h\ s i C09E FLOO3 TAl4K I
K\ FLOLS BEGINS 8 I HOT LEG SUBC00 LED g _
.C O \
~
kN 9 , 5 ' 500 5 k\gs s l 5 EN ' e is %
- @N s . ,
400 - s /~ ~ \ \ 8 s b/ \ )
@$ o t
\
KEY B Q \
\
e: HOT LEG (PRZR LOOP) 3 \' 300 - a: CORE a: COLO LEG (PRZR LOOP) ' g$
-- : SATURATION LINE $
e 9. 200 ' ' ' ' ' i 0 2 4 6 8. 10 12 Transient Tinie (liinutes)
~
Figure 3.5 _ 44 -
1 l l CGGLAt:T TE"FERATU3ES VERSUS Titt.;;SIE :T T!"E (10?# FP, BE'JINNJi<G 0F LIFE,12.2 F12 '00'J'sLE E;C P.'ET!:::E," :"IT !Cf.TED STEl.. LI!!E B..E!.Y.,P.C l FU.lP TRIP ) 1
~
703 ;
- -q PflESSURIZER EEPTIES l HOT LEG
( p t BEGINS TWD-PHASE v C00. "'s CCRE FLOOD TANK N s FLO?l CEGit:S - s, C 00b a s I HOT LEG SUBC00 LED
^
! % y U
5 500 - s
= a a A -
a f
- a O 'Ds o a Ds
- a a b s'O E 6 s E a a b%/
0 400 - 6 o
- 6 O O a O .
O
.a
& O KEY a o e: HOT LEG (PRZR LOOP) 300 -
a G: CORE a: COLD LEG (PRZR LOOP) o
--: SATURATION LINE -
200 ' ' f 1 0 2 4 6 :8 10 Transient Time (llinutes) Figure 3.6 l i i
~ , , .,
. . 1 STEA" EUL LE VOLU:.E VERSUS TP.A!.SIE!'T Til:E (102 J FP, f.EGl!";li;G OF LIFE.12.2 FI2 ggg.;tg ,
E!O I,UP1'e.1E, U::"lilU1E3 STEA:'l!!T ! . i!./..i 12 = _r 2 it il 10 -
,1 )
II '\ I \ l \ l 1 ~ 1 1
^
m 8 - I t
- I j l
\
N
= /
7 I
\
o i 6 d. - \
=
y t i
/ s l k l
G 4 - f l f% ~
\
\
'o \
\ .
g l ' r \
\
s t . I i 1
- h 1
' s :
f l i lh l /pa s i fli, 8 I Aso s% t/ I s tl I s 0 l. , 0 2 4 6 8 Transient Time (Minutes) -
- Fipre 3.7 KEY O: HOT LEG (PRZR) - RC PUMP TRIP ~
G: HOT LEG 'B' LOOP-RC PUMP TRIP 6: HOT LEG (PRZR LOOP) - NO TRIP 0: HOT LEG 'B' LOOP-N0 TRIP
l l i 1 I l i C0"E DUILET P?rs!.nFE yEPSDS TR'.*!SlHIT TIPE (102. FP., C 0!!:::!:::: 0F LIFE,12.2 FT2 C:l!"LE EhD EUFierti,U:C:lllGAIED STEA:; Lit.E CitEAK) 2500 , ,
)
2000 - W; 8 o O ' 5 1500 - s a e o KEY E
=
O: RC PUMPS TRIP o . 5 1000 _ a: NO PUMPS TRIP S 6 0 - O a O a O 500 - o o-6 a O$ o $. 0gQooa o - a a .
, o0 0 ' ' i ; i O 2 4 6 8 10 12 Transient Time (Minutes) l Figure 3.8
I
?.
3 j ta
./* -
--o-g 4
- M a
-f -> <3 / -l o
4
>=
. m e
e A Q i z- o w a -me
; w
=.a
. p G >- >J _
g e m A :7" I 8.5 2L Q & t g
= p- .to 86L uJ (;~ W
. 3 t.5 f.3 .5 r[
l to = *ZO ^^ *.
, :2: m to v m ch. .-
, ; e A A. . -
Sa t.
= w - m
! .- ce a c: .ce. *- c.,
CL == - ~ ga C.= 7"'. D p CL.
= A Sn CL.
A 2 4 r.I' E4 to = >= ."3 D D
= m LaJ A A
=3 E3. Ad n
- >. O M C OQ "
- Z D C = CE" w
J w A ce v v y"
> w o es a =
w ce = w w e :-
? , m v v > > 2 c3 c3
=3 w w ,
c -
- o o, w =
m > > m re - e o m.- w w ,o_ ,o_
> w J <- *-
l a < se w . = ce ce a= ua w w w c- a g w ~ ~ z a =
~ ~ _ w w -
ce
= w
= ce a w
= m <)
a c: e- in .a am <= m w <n <n < < , " so - m w w w +- w e ce ce . - . - . ,, Ee A Ch. 44 to A A
*, O
, c3 . . . . . . . .
j 5E- n O <3 o e sr: z = > m o_ z =m m - m W w EE' a: w =
., m =
= w w . =
w a. - w 2
= m 4 4
; w w" w $) N .
.- = .-
en - an v si G*~ 1 g 4 > l e
,.. . 1 m o l I o
,e a a a e, m -
1 1 (tant) l3A31 J71d/*S*$ l 5 Flp. tire 3.9 _ 43 _ I I
- i*
[
REFERENCES 2 E.M. Dunn, et al., "B&W's ECCS Evaluation Model," BAW-10104., Rev. 3, August 1977. 2 Intter, J.H. Taylor (B&W to S.A. Varga (NRC), July 18, 1978. 3 R.A. Hedrick, J.J. Cudlin, and R.C. Foltz, "CRAFr2 - Fortran Program for Digital Si=>1stion of a Multinode Reactor Plant During Loss-of-Coolant," EAW-10092, Rev. 2, April 1975. 4 J.F. Wilson, R.J. Grenda, and J.F. Patterson, "The Velocity of Rising Steam in a Bubbline Two-Phase Mixture," ANS Transactions, 5, (1962). _ . e e S E 49 -
l l AVAILABLE L1001D VOLUME VS TIME l FOR 0.075 FT2 BREAK IITH 1.2 ANS DECAY HEAT CURVE 3000 . 1
^
2500
)
{ -
. E E
S 2000 - 3 9' LEVEL OF ACTIVE CORE
= ________ __
1500 - 2 E
!! 1000 -
- i
'j/ RC PUMPS OFF
- 500 ' ' ' '
O 400 800 1200 1600 2000
' Time, see Figure 2-12 e
e 1 I 9 l l l -.
d 2 l RC PRESSURE VS TIME FOR 0.05 FT BREAK WITH 1.0 AND 1.2 ANS BEFORE AND AFTER PUMP TRIP , 0.05 FT2 , 2 HPI'S 1.2 ANS, PUMP DN 3000 _ 0.05 FT2 , 2 HPI'S 1.0 ANS, PUMP DN 0.05 FT2 , 2 HPI'S ! 2500 -
~ ~ ~ 1.2 ANS, PUMP OFF -l 0.05 FT2 , 2 HPI'S
.0 ANS, PUMP OFF 3m. 2000 -
a. 1500 r a- I L
\ ~
1000 - **'-7 '
~. , , -
s.% g' % 500 - k$g* ,, N 0 0 . 500 1000 . 500 2000 2500 3000 Time, sec Figure 2-13 l I
)
1
1 I I 2 PERCENT SYSTEM VOID FRACTION FOR 0.05 FT BREAK WITH 1.0 AND 1.2 ANS BEFORE AND AFTER PUMP TRIP
/
100
-- =:: .
h 80 .
#f ,~ % = . s./ /, ? /Y 60 -
l _ _ 0.05 FT , 22 HPI'S, PUMP DN, 4//([ [g : 1.2 ANS n
/ --
0.05 FT 2, 2 HPI'S, PUMP DN,
% 40 -
' l.0 ANS y f
= / 2
*- / -o _0.05 FT , 2 HPl *S. PUMP OFF, 1.2 ANS
/
20 - 2
- z _ 0.05 FT , 2 HPI'S, PUMP OFF, 1.0 ANS 0 i ,
0 400 800 1200 1600 2000 2400 2800 Time, sec Figure 2-14 l l
l l AVERAGE SYSTEM V010 FRACTION VS TIME FOR A'0.075 FT 2 l BREAK, BREAK LOCATION COMPARISON PUMPS OFF e 905 V010 i 100 UNC0VERY TIME = 625
* ~
,/ SEC
- 80 -
/ s Y /h -
\
j f'& UNC0VERY TIME =
. 60 -
,/ g 450 SEC
/ W4 E" /A
/ *
- 40 -
5 /
~
' 9. -
f E 20 - f/
' ' ' ' ' l 0
0 400 800 1200 1600 2000 2400 Time, see l i Figure 2-15 1
- ,, , - -~- -- w-
- l COMPARISON OF del.IVERED HIGH PRESSURE INJECTION FLUl0 TO RV FOR PUMP OISCHARGE BREAK 1200
\
-N, 1100 -
Nf 1000 - N# 900
\/
N ' 1 % 3 B00 - g ~ g' 700 -
\
600 - \ '# 2 # '0
\ 0s #
j 500 -
\ #p,
,5 400 - \
ED zgg, 300 - % 200 . I00P I yp, N< j 100 - 0 ' ' ' ' ' ' ' I ' ' - ' I O 200 600 1000 1400 1800 2200 2600 3000 I Pressure, (psia) Figure 2-16 G e b l
l 1 O w i W X '
.n
( e--D< i
.f.-.----:., .
50
.___________-..________.,l5 Q .
@ 23. l p _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .-
6.......s s MAcv b -- I
' @ I h $ E010Y i b *__
*T*
" -1 at :-@ If . ,-- 1 ut e- t '
. .2 :
l _eJ.~ M 18 5 ja l O @ l ' _O q fo - l 6(. 4 . J -
@[ 1G c @
). . M d7 12, v
@ 4 11 i O~ 3 - /6 i O 1 S[ -
$[ @%
"e
'. . 2s 'S 'n .. j , a l l A
. 6 -
?g &h -
. LT If L t 3
g' @ N pg 'O O $ Figure 3.1 ! MINITRAP2 Noding and Flow Path Schmaa 1 w
l'iiESSU.ilZER A ] STEA!: Cd: ERA 1CR LIC'J10 LEVEL VEFSUS T!',*.!:SIENT Tl"E 2 ' (102 FP. Ei:3 0F LIFE,0.6 FT STEA': Lit:E C..EAK (00'J::0!!:G CCDERATE FREU. ). (RC PU:.'P TRIP) S.G. 10BE REGl0ft ooeogo 50 - FULL 0 C o g-o - O G a w 40 - g PRZR. FULL a U O g O e o
*- O 0 U 5 O o
g 30 - g g g e 6 g se
" 8e U l 3 eoo 0 U
3= 20 'l G O G S 5 G
- a e
- 6 KEY Z G 10 -
6 0: PRESSURIZER 0 6 o: STEAM GENERATOR 'A' a: STEAM GENERATOR 'B' h i i i , , i -i 2 4 6 8 10 12 14 16 0 f isient Time (Minutes)" , 1 Figure 3.2 1
. .n-
- s Ca w
- z
< _m
=
- = ce ca o D""
e== >= x e 4 c W z s CE tc ty: GE d\
= \
$50
-m m z x w ce w w - <
4 = - ex e> s = => m .:. _-. .
.- cu w
- t.1 ~
ce s w < e w w w 3 = % sg a. g cc e- ,-
' a m -
25 w ce
\s ,
s s s - r = E- N, s a a a w - o e.c - o w z s s w a
> - s s = c w _, n. s, ss s --
e :2 - 's g g .s. 4 ce g - cm w - s s g E G o_ ' g h - e - o
.E 2, m.- E z \
o \s ss .- te w o. o L*J s s
\
\
e c o cc w s s e- m. ce s \ _u
- o m-- a s se s ME DN - Mm *o, w-TM - = g 5=w - =w \
w - o .- s mw-
-2 g .m w s .
s --w
= w a o.
= N e n. m \ss .
< ca ca . N a w cn gg g w a w D- a -
Ds - o w z - c3 2 ^ ss g N ss z z gs
= w m
e g % cc w ce g i g w m w 5
= a.
-w N b 6 - ,
a w a m sn
= w w w ~ o ac = a G4 O
- n. -
w
=
B 4 0 - m i G 4 O O GG r . G
- i i , e, , o
- .J
= = =, a .
= w < m _
(3 D IDAD1 DlHhl1 J01CJau 3 mCalg,*J32lJnss3Jd Figure 3.3 ee
.% w - , -
1 1 i e h
. oo c -
11
^
h 00 q - m w a I w a u~ I oo c g 9 y
- ' ee o m =
& s a ~_
n , x m g om e ~
=
m a a.
~ -
m -x ;
-o a oa e a.
v o m a 8 ga m - z
- a g
4 a w ~
" oo c "i E m x 5,
= I om n w o
.- a a o- - o m
=
e I e o a < - o - = " " " E ggS oo q = m - 1 -
> a =. oO G . . . . . . . .
m E u g oc 4 eoo i 5 m m o i >. 00 o M - EGE l 8 O~~ ee o m w m t +e w u. m o g oe 4 e a. I es 4 m 5 *. 5 l
~
l om e . _
= E. .
.- a. - !
z
- u. I oa q - -
1 l 4 gM 0e ! l EEv I 00 4 I oO G _ I e o I e c I y CD C ED 4 - . I ED 4 1 cp o i
/
f Cao
'd c t ,
8 g g g : o o
=
- = = ,
)
(.4.) aJnteJadmal suctoo3 Figuie 3.4 e
l
. Figura 2-1. CRAFT 2 N: ding Diagram Icr smaa.J. arcas, l l
T' O ~, = -- 34 13 -
' ~
j ;; - , - 4 ,. ..
~
, g3 if (i) As s
,b is 6 <s
,. O G i. BJ 'Ef [3 (HF e g@ al 3e-
, ,. u E,-
=
6 - i i Q , 8g - . ... . Ia gg Bud. 22 1s C 8e 3.s.
" u O n, - GG O ...,
e m..,... ,. 7 , ca) - . . - . . i e = 7 e ..., 6 ........ m. ... ....
.. n>> ........ < ... .
v se.., sr..= pode No. Tdentifiesticm Path Wo. Ydentiffeattem 1 Douscomer 2,2 Core 2 Imver Plenum 3,4.18.19 . Bot Les Piping.
. 3 Core. Core 3ypass. Upper S.10 Eat 34g. Upper Plenum. Upper Read 6.21 SC Tubes -
4.14 Not 1,eg Piping 7.22 s p er Head S,15
- Steam Generator Upper 3 Core 3ypass Bead. SC Tubes (Upper Half) 9.13.24 Cold Leg F1 ping 6.16 SC Tubes (Lower Half) 20,14.25 Fusps 3.12 SC Lower Bead 21,12.15.16,26.27 Cold 14g Piping
^
9.11.19 Cold Les Piping (Fump Svetion) 17,31 Devocomer. Cold Les Piping (Fuop Discharge) 23 ' 3,71 10.12.20 .
. 13 Dnper Downconer 28,29 Upper Downco=ar
*(Above the q,of Nozzle Balt) 30 Pressurizer 21 Pressuriser 32 Yeat Ysive 22 Containne t 33,34 Laak & Return Fath 33,36 NF1
! 37 Containment sprays l l
P1::ure 2-2. CRAFT 2 N0 DING DIAGRAM FOR SMALL BREAKS (6.N0DE MODED CFT 11
@ 3 2 6 5
._ N- 1 --
l 1 l e s % ~@
- ~, g ,
n@ k LEAK PATHS 8 & 9 M@ ( H as Wode No. Identification Path No. Identification 1 PD Piping. DC, LP 1 Con 2 Primary SG 2 LPI 3 Core, UP, Hot Legs 3,10,11 BPI 4 Pressurizar 4 Not Legs 5 Containment 5 Pinups 6 Secondary SG 6 Vent valve
- 7 Pressurizer 8,9 Leak & 1sturn Path l
l ~ 6
J CORE P" ESSURE VS TIME,177-LL, 2772 Ht, PUMPS ON 2
~
0.025 FT BREAK 23 NODE MODEL 20 0.025 FT 2 BREAK
~ 6 N00E N00EL 18 S -
.O '
.! 16 2.
a a 3 14 E a i 12 - 10 - g , , 500' 1000 1500 2000 2500 0
' Time, sec ,
Figure 2-3 O
a l PERCENT SYSTEM VOIDS VS TINE, PUMPS ON 100 80 - m 2 i* SD - ( ess # .sas" B 9.DD ' DD a.
- ,, " T (B %
t .0D 2D -
.+-
, f'$
r - I
' . i , ,
a . 0 400 800 1200 1600 2000 2400 Time, sec Figure 2-4 , O e i 9
- - 27 -
- w m w
BREAK SPECTRUM-RC PRESSURE flTH THE RC PUMPS OPERATIVE AND 2 HPI PUMPS 2500 2000 - i 1500 - M Y l 5 .
\
p N 'N 1000 - .
-. ~ . ' * % . ~ 0.025 FT 2
\. N s x %~~ ~ s
~
\.. Idb ' -0.05 FT2
- x --- e _ 0.075 FT2 2
\ .~ 0.2' FT2 0.10 FT 0 ' ' ' '
O 500 1000 1500 2000 2500 3000 3500 Time, sec ) l Figure 2 5 9 o
,,ww w e--., -
-,y -- ,- ---r---- y---wwy_
BREAK SPECTRUM-AVERAGE SYSTEM VQl0 FRACTION WITH THE RC PUMPS OPERATIVE ANO 2 HPl PUMPS 100 p "~"- /yo.go p, am, P g # ,
- J- #
- / *
* / ~
* *~
~,s..',,
e 80
-ql '* : R ojQ/ s j/ /
i :l %;/9+/ ,- *. o% %/
- b/
& SD ~! \
m : o l +/ &
; ia/ /v/ "
40 i: ) /+ /
= -:'/
- o / /
/
gE .*/./* E
! /
o +/ l s9L - 3 20 /
/.
/.s- '
l' ' 0 ' ' ' ' ' O 400 800 1200 1600 2000 2400 2800 Time, sec
. Figure 2-6 e
9 O e 1
O.1 FT 2 BREAK WITH CONTINUGUS RC PUNP s OPERATION AND 2 HPI PUMPS 100 2500
,r~~~~ __,._,_ _ _
LPI
/
- k
/
/ 's\% -
2000 80 -
/ I\%
m j
. % \
y I , h 60 - I
\g - 1500 g-h 4 5> l s I
= 5
%.J.,.%s, - 1000 ;
l g 40 - i _
=
3 i s 3 i N
- I N LPI _ 500 i
20 - l
\*%~~ % - . , ,
0 ,
, , , , s '~ ~ - 0 -
500 1000 1500 2000 2500 3000 0 Time, sec
* ~
Figure 2-7 l 9
- 30 -
a wy-- , , , . _ , , , _ , _ _ , _ _ , _ _ _ _ _ _ _
* - - - r
RC PRESSURE FOR 0.05 FT2 BREAK AVAILABLE 1 HPI VS 2 HPI'S 30 .
---- 2 HPI'S, PUNP DN, HONOGENEOUS 25 -
- - H1 UMP DN,HONOGENE005
#=
20 - M V 2 E- 15 i a !
\
= s_ --. wo~s% ,%
10 - N
% No N
ss N *0% %'ON o 5 N-Q 0 , , , , , 0 500 1000 1500 2000 2500 3000 Time, sec Figure 2-8 4 e e
2 AVERAGE SYSTEM VOID FRACTION FDR 0.05 FT AVAll.ABl.E 1 HPl VS 2 HPI'S 100 - -
.-o_.
. ,/g# --
80 -
/ ,/,,,
m /
. / -
g
- 60 -
. l//
---2 HPI'S, PUNP DN, HOMOGENEOUS a / * - 1 HPI, PUMP ON, HOMOGENEOUS
' }
5 *t 5 / Z 40 -
./ s' 3
' /
/
/
20 l 1 0 . , , , , , , 0 400 800 1200 1600 2000 2400 2000 3200 Time, sec Figure 2-9 l
\
. l e
e e _ - p- , - ,-
1 2 RC PRESSURE FOR 0'.075 FT , PUMPS OFF e 905 SYSTEM VOIO 30 _ _ _ _ 2 HPI'S, PUMP DN, 25 - HOMOGENEOUS l
--- 2 HPI'S, PUMP OFF @ ;
905 VOIO, 2 PHASE 20 - -!l O E 15 - J E k s__ r E 10 - __'s s s\t .
'Ns 5 -
. q,s* -
N,%.g , * ** * ===. . . . . 0 O 500 1000 1500 2000 2500 3000 Time, sec' Figure 2-10 , 1
AVERAGE SYSTEM V010 FRACTION FOR 0.075 FT ,2PUMPS OFF e 905 SYSTEM V010 100 f-------_,,,,,_
,sA ~~.
80 -
/
,N.%
* / .
a f o / 60 - / f, / I - m /
,/ ---- 2 HPI'S, PUMP DN HOMOGENEOUS 3 40 -
,/ -.- 2 HPI'S, PUMP OFF e 905 V010, 2' / 2 PHASE
/
20 - 0 O 400 800 1200 1600 2000 *2400 3200 Time, see Figura 2 11 e J e I e 34 - y -
. . - _ _ _ ,.-,.c-. .% ., _w, me. .
i l tion case refilling about one minute before the case with i two of four pumps running '(See Figures 3.2,3.3). In both cases, the system is highly subceoled, from a minimum of 30*F to 120*F '
)
and increasing at the end of 14 minutes (refer to Figure 3.4). i i
~
- It is concluded that an RC pump trip following HPI actuation will not increase the probability of causing a LOCA through i
the pressurizer code safeties, and that the operator will have the same lead time, as well as a large margin of subcooling, to control HPI prior to safety valve tapping. Although no case
- with all K:: pumps was made, it can be inferred from the one loop case (with pumps running) that the subcooled margin will be slightly larger for the all pump's running case. The pressurizar will take longer to fill but should do so by 16 -{
minutes into the transient. Figure 3.tshows the coolant temperatures (hot leg, cold leg, and core) as a function of time for the no, RC pumps case. .
- 2. Effect of Steam Bubb'le on Natura Circulation Cooling For this concern, an analysis was performed for the same i ,
generic 177 FA plant as outlined in Part 1, but assuming that as a result of an unmitigated large SLB (12.2 ft. DER), the s excessive cooldown would produce void formation in the primary i ~ system. The intent of the analysis was to also show the extent of the void formation and where it occurred. As in the case analyzed in Part 1, the break was symmetric'to both generators such that both would blow down equally, marimizing the cooldown (in this case there was a 6.1 ft.2 break on each loop). There was no MSIV. closure during the transient on , either steam generator to maximize cooldown. Also,, the tur-i bine bypass system was assumed to operate, upon rupture, . until isolation on ESFAS. ESFAS was initiated on low RC pressure and also actuate 1 HPI (bofh pumps), tripped RC passps (when applicable) and isolatb the MFWIV's. The AFW was initiated to both generators on the low SG pressure signal, with minimum delay time (both pumps op'erating). This _ analysis was performed twice, once assuming all RC I pumps running, once with si; pumps being tripped on the IIPI actuation (after ESFAS), with a short (%5 second) delay. In both cases, voids were formed in the hot legs, but the dura-i . . , ( . l l .. . . . , _ ,
L . . . tion and size were smaller for the case with no RC pump trip (refer to Figure 3.7).Although the RC pump operating case had a higher cooldown rate, there was less void forma-tion, resulting from the additional system mixing. The coolant temperatures in the pressurizer loop hot and cold legs, and the core, are shown for both ' cases in Figures 3.5, l 3.6. The core outlet pressure and SG and pressurizer I levels versus time are given for both cases in Figures 3.8, 3.9. This analysis shows that the system behaves very similarly with and without pumps, although maintaining RC pump flow does seem to help mitigate void formation. The pump flow case shows a shorter time to the start of pres-surizer refill than the natural circulation case (Figure 3.9), although the time difference does not seem to be very large.
- 3. Effect of Return to Power There was no return to power exhibited by any of the BOL -
cases analyzed above". Previous analysis experience (ref. Midland FSAR, Section ISD) has shown that a RC pump tr.dp will mitigate the consequences of an EOL return to power condition by reducing the cooldown of the primary system. The reduced cooldown substantially increases the suberitical margin which, in turn, reduces or eliminates return to power. , D. Conclusions and Summary A general assessment of Chapter 15 non-LOCA events identified three areas that warranted further investigation for impact of a RC pump trip on ESFAS low RC pressure signal.
- 1. It was found that a pump trip does not significantly shorten the time to filling of the pressurizer and approximately the same time interval for operator action exists.
- 2. For the maximum overcooling case analyzed, the RC pump trip increased the amount of two-phase in the primary loop; however,thepercentvoidformatiodisstilltoosmallto
. affect the ability to cool on natural circulation.
- 3. The subcritical return-to-power condition is alleviated by the RC pump trip ease due to the reduced overcooling effect.
Based upon the above assessment and analysia, it is con- I cluded that the consequences of Chapter 15 non-LOCA events are not g . . e =
Encreated dua to tha cdditica cf a RC pump trip on ESFAS low RC pressure signal, for all 177 FA lowered loop plants. Although there were no specific analyses performed for TECO, the conclusions drawn from the anzlyses for the lowered loop plants are applicable. , l
- i f
w 4 e 4 e e e l e 0 0 9' O l l 9
l Table 2-1. Analysis Scope With A N Available Continuous RC Break locati n pump parati n RC pump trip 3 90% void
. Break size,
_ jft 2) 2 HPI 2 DI Cold leg Hot leg 1 DI - 0.025 I I 0.05 I I* I I* 0.075 I I I I 0.10 I I I 0.20 I I Analyzed with both 1.0 and 1.2 ANS decay curves. O e 4 l 1 3 - .v v .r
J Table 2-2. Impact Assessment of Break Spectrum With RC Pump Trip at 90% void Break size (ft2) Core uncovery time (sec) 0.10 550 0.075 625 0.05 575 Notes: 1. Two HPIs available during the transient.
- 2. Core uncovery time is the time period following pump trip re- _.
quired to fill the inner RV with water to an elevation of
- 9. ft in the core which is ap-proximately 12.ft when swelled.
4
~
21 - a r-.- - -m , - ---------,-e. , , - , - - , , , , , , - . , , , , - - , , - - ,
Table 2-3. Comparison of System Void Fractions at ESFAS Signal System void fraction ,
* ' ^ '
Break size, - (ft 2) Pumps on Pumps tripped 0.02463 0.0 0.04 4.47 0.05 0.04 0.055 6.74 0.07 8.06 0.075 0.90 _. 0.085 8.45 1 0.10 2.17 7.97 i i 0.15 10.70 O.20 6.78 s . e I e
, - - -,m-~ ,- ,.. - - -- - - - - - -- - - - < -
MINITRAP2 NODE DESCRIPTION __ EDDE NUMBER DESCRIPTION 1,33 Reactor Yeasel, Lower Plenum 2,34 Reactor Vessel, Core
. 3,35 Reactor Vessel, Upper Plenum 4,10 Hot Leg Piping '
5-7,21-13 Priautry, Steam Generator
. 8,14 Cold Leg Piping 9,32 Reactor Vessel Downcomer 15 Pressurizer 16,24 Steam Generator Downcomer 17,25 Steam Generator Lower Plenum 18-20,26-28 Secondary, Steam Generator 21,29 , Steam Risers 22,30 Main Steam Piping 23 Turbine 31 Containaent _
MINITRAP2 PATH DESCRIPTION l PATH NDIBER DESCRIPTION
)
1,2 Core 45,46 Core Bypass i 3,5,5,11,12.44 - Bot Leg Piping ' 6,7,13,14 Primary, steam Generator
- 8.15 RC Pumps 9,16 Cold Leg Piping 10,43 Downcomer, Reactor Vessel 17 Pressurizer Surge Line 18,19,26,27 Steam Generator Downcomer 20,21,28,29 Secondary, Steam Generator 22,30 Aspirator 23,31 Steam Riser 24,32 Steam Piping 25,33 Turbine Piping .
34,35 Break (or Leak) Path 36,37 BPI 38,39,43,44 APW 40,41 Main Feed Pumps . 42 LPI : Table 3.1 l
ATTACHMENT 1 DUKE POWER COMPANY l OCONEE NUCLEAR STATION ' Response to IE Bulletin 79-05C l SHORT TERM ACTIONS Item 1 On July 30, 1979, the following actions were taken: A. The appropriate Emergency Procedures, EP/0/A/1800/4, Loss of Reactor Coolant and EP/0/A/1800/8, Steam System Leak-Rupture, were revised to - require tha.: upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, all operating reactor coolant pumps are tripped. B. A letter was sent by the Superintendent of Operations to all Shif t Supervisors requiring all Control Rooms to be manned as described in Bulletin Item IB. Item 2 Attachment 2 " Analysis Summary in Support of an Early RC Pump Trip" is pro-vided in response to this item. Section 3 of this report, " Impact Assessment of a RC Pump Trip on Non-LOCA Events," is included to allow the developmaat of preliminary non-LOCA guidelines as required by Item 3. Item 3 1 Attachment 3, " Guidelines for Operator Action," provides guidslines on LOCA and non-LOCA transients. Iteu 4 Following review and comment by B&W utilities, formal operating guidelines will ca issued. At that time, appropriate' emergency procedures will be revised and all licensed reactor operators and senior reactor operators on shift will be tra$ned 1979. on the new guidelines. This will be completed on or before September 7, Item 5 A preliminary response is provided in Attachment 2, Guidelines for Operator Action, Section III, " Criteria for RCP Restart." However, additional analysis is required and will be completed by October 31, 1979.
_2_ r LONG TERM ACTION Item 1 Attachment 4 provides a schematic of a conceptual design which would provide automatic tripping of operating reactor coolant pumps (RCP's) on coincident low reactor coolant system (RCS) pressure and low RCP power. The design would use existing engineered safety feature (ESP) low RCS pressure and reactor pro-tactive system (RPS) RCP power monitoring signals for input to an 'and' gate. Output would be to either the RCP breakers or the normal or startup feeder breakers. The pwar sources of each channel are completely independent. This RCP trip could be installed the first refueling outage for each unit following six months after NRC approval. 1 6 I
4 1-l ATTACHMENT 2 - ANALYSIS
SUMMARY
IN SUPPORT OF AN EARLY RC PUMP TRIP
, , g. ,. _ _ . . . _ _ . - -
. m ,
CONTENTS . Page I. INTRODUCTION . . . ............. .. . . . . . . . . .. 1 II. SMALL BREAK ANALYSIS . . . . . . . . . .. .... . . . . . . . 2 A. Introduction . . . . . . . . ... ......... . . . . . 2 B. System Response With RC Pumps Running ... . . . . . . . . . 2 C. Analysis Applicability to Davis-Besse 1 ... . . . . . . . . 11 D. Effect of Prompt RC Pump Trip on Low Pressure ESFAS Signal . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E. Conclusions ............... . . . . . . . . . . 13 o III. IMPACT ASSESSMENT OF A RC PUMP TRIP ON NON-LOCA EVEh3 . . . . . . 15 A. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . 15 B. General Assessment of Pump Trip in Non-LOCA Events . . . . . . 15 , C. Analysis of Concerns and Results . . . . . . . . . . . . . . . 16
, D. Conclusions and Summary .. ... ..... . . . . . . . . . 18 I
e { 4 I O I \ l 1 l _ _ ., _, , _ , - - , - - - - , - ~ * -
l , s -- s ANALYSIS
SUMMARY
IN SUPPORT OF - AN EARLY RC PUMP TRIP
-l I. INTRODUCTION _
BW has evaluated the effect of a delayed RC pump trip during the course of small loss-of-coolant accidents and has found that an early trip of the RC pumps is required to show conformance to 10CFR50.46. A sununary of the IDCA analyses performed to date is provided in Section II. This discussion includes:
- 1. A description of the models uti.tzed.
- 2. Break spectrum results with continuous RC Pump Operation.
- 3. Break spectrum results with delayed RC pump trips including estimates of peak cladding temperatures.
- 4. Justification that a prompt pump trip following ESFAS actuation on low RC pressure provides LOCA mitigation. ,
An impact assessment of the required pump trip on non-LOCA events has also l been completed and is presented in Section III. This evaluation supports the use of a pump trip following ESFAS actuation for LOCA mitigation since no detrimental consequences on non-LOCA events were identified.
. _ , . ._.7 ,
II. SMALL BREAK ANALYSES . A. Introduction . Previous small break analyses have been performed assuming a loss-of-offsite power (reactor coolant pump coastdown) coincident with re-l , actor trip. These analyses support the conclusion that an early RC pump trip for a LOCA is a safe condition. However, a concern has been identified regarding the consequences of a small break transient ir. , which the RC pumps remain operative for some time period and then are lost by some means (operator action, loss-of-offsite power, equipment failure, etc.). This section contains the results of a study to further understand how the small break LOCA trans h t evolves with the RC pumps _, operative. Specifically, section B. des u bes the system response with the RC pumps running for B&W's 177-FA lowered-loop plants. In-cluded in this section is the development of the model used for the analysis, a break spectrum sensitivity study, and peak cladding tem-perature assessments for cases where the RC pumps trip at the worst time. Section .C ' demonstrates the applicability of the conclusions
~ drawn in section. 3 to a 177-FA raised-loop plant (Davis-Bessa 1).
The effect of a prompt tripping of the RC pumps upon receipt of a low pressure ESFAS signal is liccussed in section .f3.1 Finally, sec-tion.E ' summarizes the conclusions of this analysis. B. System Response With RC Pumps Running
- 1. Introduction Recent evaluations have been performed to *==4ne the primary system response during small breaks with the EC pumps operative.
During the transient with the RC pumps available, the forced , circulation of reactor coolant will maintain the core at or near the saturated fluid temperature. However, for a range of break sizes,, the reactor coolant system (RCS) will eveN e to high void fractions due to the slow system depressurization and the high liquid (low quality fluid) discharge through the break cs a re-
, 'sult of the forced circulation. In fact, the RCS void fraction -
will increase to a value in excess of 90% in the short term. In i
i the long term, the system void fraction will decrease as the RCS RCS depressurizes, HPI flow incrasses, and decay heat diminishes. With the RCS at a high void fraction, if all RC pumps are postu-
, l I
lated to. trip, the forced circulation will no longer be available 1 and the residual liquid would not be sufficient to keep the core covered. A cladding temperature excursion would ensue until core i cooling is reestablished by the ECC systems. The following para- j graphs summarizes the results of the analyses which were performed l for the 177-FA lowered-loop plants, to develop the consequences of this transient. I
- 2. Method of Analvsis -
The analysis method used for this evaluation is basically that de-scribed in section 5 of RAW-10104. Rev. 3, "B&W's ECCS Evaluation Model"1 and the letter J.H. Taylor (B&W) to S.A. Varga (NRC), dated
- July 18, 19782 , which is applicable to the 177-FA lowered-loop plants for power livels up to 2772 MWt. The analysis uses the CRAFI23 code to deve'op the history of the RCS hydrodynamics.
1
+
However, the CRAFT 2 monel, used for this study is a modification of the small break evaluation model described in the above ref-erences. Figure 2-1 shows the CRAFT 2 noding diagram for small l breaks from the above referenced letter. The modified CRAFI2 model consists of 4 nodes to simulate the primary side,1 node for the secondary side of the steam generator, and 1 node representing the reactor building. Figure 2-2 shows a schematic diagram of this model. Node 1 contains the cold leg pump discharge piping, downconer, and lower plenum. Node 2 is the primary side of the SG and the pump suction piping. Node 3 contains the core, upper ple-num, and the hot legs. Node 4 is the pressurizer and nodes 5 and
- 6 represent the reactor building and the SG secondary side, re-spectively. This 6 node model is highly simplified compared to those utilized in past ECCS snalyses. It does, however, maintain RCS volume and elevation relationships which are important to properly evaluate the system response during a small break with the RC pumps running.
l l
The breaks analyzed in this section are assumed to be located in the cold leg piping between the reactor coolant pump discharge and
, the reactor vessel. Section B.7 demonstrates that this is the vorst break location. Key assumptions which differ from those de-scribed in the July 18, 1978, letter are those concerning the equip-ment availability and phase separation. These are discussed below.
- a. Equipment Availability The analyses which were performed assumed that the RC pumps re-main operative after the reactor trips. For select cases, after the system has evolved to high void fractions (approxi-
! mately 90%) the RC pumps were assumed to trip. Also, the in- - pact of 1 versus 2 HPI systems for pump injection were anafned.
'lhe majority of the analyses performed assumed 2 HPI pumps.
However, as is demonstrated later, even with 2 HPI pumps avail-able, cladding ta..peratures will exceed the criteria ef 10 CFR 50.46 using Appendix K evaluation techniques. Therefore, fur-ther analysis with only 1 HPI pump would only be academic.
- b. ,P_hase Separation The present ECCS evaluation model created to evaluate
~
small breaks without RC pumps operative,(quiescent RCS) uti-li=es the Wilson" bubble rise correlation for all primary sys-tem control volumes in the CRAFT evaluation. In this analysis, for the time period that the RC pumps are operative, the pri-me y system coolant is assumed to be homogeneous, i.e., no phase separation in the system. In reality, the flow rates in the core and hot legs are low enough that slip will occur. This will cause an increased liquid inventory in the reactor vesse2 compared to that calculated with the homogeneous model.
- With the homogeneous assumption, core fluid is continuously circulated thwnthout the primary system and a portion of that fluid is lost via O s break. During the later sta$as of the transient, a slip model will result in fluid being trapped in
, the reactor vessel and the hot legs. The only method of, losing liquid during this period will be by boiling caused 'y the core decay heat. Thus, the assumption of homogeniaty for the period I with the RC pumps operative is conservative.
_, , - . ~ . , . _ .
Following tripping of the RC pumps and the subsequent loss-of-forced circulation, the system will collapse and separate. The residual liquid will then collect in the reactor vessel and the loop seal in the cold leg suction piping? For this period of the transient, the Wilson bubble rise model is utilized. The homogeneous assumption for the period with the RC pumps
! operating applies to nodes 1, 2, and 3 in the CRAFT model.
Node 4, the pressurizer, and node 6, the secondary side of the steam generators, utilize the Wilson bubble rise model throughout the transient as these nodes are not in the direct path of the forced circulation. __
- 3. Benchmarking of the 6 Node CRAFT Model Studies were performed to compara the results of the 6 node model to the more extensive evaluation model for B&W's 177-FA lowered-
- loop plants as described in the letter J.H. 'Esylor (B&W) to S. A.
3 Varga (NRC), dated July 18, 1978. The break size selected for I this comparison is a 0.025 ft2 break at pump discharge. This break represents the largest single-ended rupture of a high energy line (2-1/2 inch sch 160 pipe) on the operating plants. The
. break can be viewed as " realistic" or the worst thdt would be ex-pected on a real plant. Figures 2-3 and 2-4 are the results of this comparison. System pressure and percent void fraction shown in Figures 2-3 and 2-4, respectively, compara very well with those from the more extensive (23 nodes) CRAFT 2 small break model. ..s seen in these figures, the difference is not significant and is less than a few percent. The computer time for this 6 node model is, however, significantly decreased. The medal utilized for this study is thus justified based on comparison of results to '
the more extensive small break model and de=1rable because of its economical run time. I
- 4. Analysis Results The break sizes awamined for this analysis remged from 0.025 ft2 to 0.2 ft2 in area sad are located in the p g discharge piping.
Breaks of this size do not r'esult in a rapid system depressuri-sation and rely pred minantly upon the HPIs for mitigation. l lj . . =
, Table 2-1 summarizes the analyses performed for this evaluatien.
The majority of the analyses performed utilized 2 HPI pumps through-out the transient. The effect of utilizing 1 HPI pump is discussed in this section. ' Figures 2-5 and 2-6 show the system pressure and average system void fraction transients for the break spectrum analyzed assuming continuous RC pump operation and 2 HPI's available. In Figure 2-6, the average system void fraction is defined as Vg-V2 Average system void, % = x 100 91 V3 = total primary liquid volume excluding the pressuri-zer at time = 0, V2
- total primary liquid volume excluding the pressuri-zer at time = t.
- This parameter was utilized in place of the mixture height in that the coolant will tend to be homogeneously mixed with the RC pumps operative. Under these assumptions, the core.is cooled by forced circulation of two-phase fluid and not by pool boiling as in the case where the RC pumps are not running and separation of steam l and water occurs. As shown in Figure 2-5, the system pressure re- l
, sponse is basically independent of break size during the first several hundred seconds into the transient. This occurs because the forced circulation of reactor coolant maintains adequate heat transfer in the steam generators; the primary system thus depres- 1 1
surizes to a pressure (about 1100 psia) corresponding to the sec-ondary control pressure (i.e., set pressure of SG safety relief valves). After seme time (250 seconds for the 0.1 ft2 break), the
- system pressure will decrease as the break alone relieves the core energy.
Figure 2-6 shows the evolution of the system void fraction; values in excess of 90% are predicted very early (300 seconds) into the I \ transient. For the larger breaks the system high void fractions occur early in time. For the smaller breaks it takes in the order of hours before the system evolves to high void fraction. Core cooling is maintained during a small break with continuous RC pump _6-
~
--mm.. *-.
- operation regardless of void fraction. In the long term, the sys-tea will depressurize and the enhanced performance of the ECCS (HPI and LPI) will result in reduced system void fraction. 1
)
)
Figure 2-7 illustrates this long term system behavior for a 0.10 ft2 l break. For this case, the LPIS are operative at approximately 2300 1 - seconds, and a substantial decrease in system void fraction results. An arbitrary pump trip after approximately 2700 seconds would not result it. core uncovery. The potential for core uncovery due to an RC punp trip is thus limited to a discrete time period during which thes natural evolution of the system produces high void frac-tions ard prior to LPI actuation. For a 0.1 ft2 break, this time period is on the order of 2000 seconds. For smaller, breaks, this - critical time could be a few hours even if the operator initiated a controlled cooldpa and system depressurization as recommended in the small break guidelines. Although the analyses described above used 2 HPI pumps, the effect of only 1 HPI pump available on the system void fraction evolution while the RC pumps are operating is not significant. Figures 2-8 and 2-9 show the impact of one versus two HPI pumps on system pres-sure and average void fraction transients for a 0.05 ft2 break with
. the RC pumps operative. As seen from these figures, the results with one HPI pump are not significantjy different to the two HPI pump case and are bounded by the spectrum approach utilized. With I one HPI pump, the system does depressudze more slonly (less steam condensation) and a higher short term equilibrium void fraction is achieved. Also, recovery of the cort following a loss of the RC pumps w>uld be significantly longer vith only 1 HPI pump avail-able. .
l I The majority of the analyses provided in this report uses two HPI j pumps and demonstrates a core cooling problem with worst time pump trip given that assumption. As analysis of one HPI available cases would only show a larger problem, such cases have not been exten-sively considered. As demonstrated in section B.4, the resolution of this problem, forced early pump trip, provides assurance of core cooling for both one or two HPIs available cases. Therefore, l l I
C. ore is no need for further pursuit'of the single HPI available case.
'1he effect of the RCP tripping during the transient was studied by , assuming that the pumps are lost when the system. reaches 90% void fraction. Loss of the RC pumps at this void fraction is expected to produce essentially the highest peak cladding temperature.
After the RC pumps are tripped, the fluid in the RCS separates and liquid fans to the lowest regions, i.e., the lower plenum of the RV and the pump suction piping. At 90% void fraction, the core will be totaHy uncovered foHowing the RC pump trip. Thus, the time required to recover the cora is longer than that for RC pump trips initiated at lower system void fractions. System void frac- - tions in excess of 90% can possibly result in slightly higher tem-peratures due to the longer core refill times that may occur. However, the peak cladding temperature results are not expected - to be significantly different as the system pressure and core de-cay heat, at the time that a higher void fraction is reached, will be lower. Table 2-2 shows the core uncovery time for the cases analyzed sith the RC pumps tripping at 90% void fraction with 2 HPI pumps avail- {
, able for core recovery. As shown, the core will be uncovered for '
appraw4==tely 600 seconds for the breaks analyzed. Figures 2-10 i sad 2-11 show the system pressure and void fraction response for
]
the 0.075 ft2 break with a RC pump trip at 90% void fractica. As seen in these figures, the system depressurizes faster after the RC pump trip, due to the change in leak quality, and t.he vaid fraction decreases indicating that the core is being refilled. ] Figure 2-12 shows the core liquid level response following the RC pump trip. The core is refilled to the 9 foot level with collapsed liquid approximately 625 seconds after the assumed pump trip. Once the core liquid level reaches the 9 foot elevation, the core is expected to be covered by a two-phase mixture and the cladding l
. temperature excursion would be terminated.
O e . I -=.---..'. .. -
. - , . _ . .. ~ . . . , , _. - - ~ _ _- . , _ , . -, m. .,. . , _ .
l i
- 5. Effect of 1.0 ANS versus 1.2 ANS Decay Curve An analysis was performed using the more realistic 1.0 ANS decay curve instead of 1.2 ANS decay curve. The study was done for a
- 0.05 ft2 break with 2 HPI;s available and pumps tripped at 90%
4 system void fraction. Figures 2-13 and 2-14 show a comparison of system p" essure and average system void fraction for 1.0 and 1.2 ANS decay curves. As seen in Figure 2-13, the system pressure for 1.0 ANS case begins to drop from saturation pressure (%1100 psia) about 200 seconds earlier than the case with 1.2 ANS as a result of reduced decay heat. Also, the system will evolve to a lower average void fraction as shown in Figure 2-14. After the pumps trip at 90% system void fraction, the case with 1.0 ANS decay __ curve has a shorter core uncovery time by approximately 200 sec-onds compared to 1.2 ANS case. This case demonstrates that the effect of a delayed RC pump trip may be acceptable when viewed l realistically. A peak cladding temperature assessment for this ' case will be provided in a supplementary response planned for l September 15th, to the I&E Bulletin 7905-C. j
- 6. Effect of No Auxiliary Feedvater Analyses have also been performed with the RC pumpg available and no auxiliary feedvater. These analyses all assumed 2 HPI pumps were available. The system void fraction evolutions for these calculations were not significantly different from those discussed with auxiliary feedwater. Thus the conclusions of the cases with auxiliary feedwater apply.
l
.2 Break location Sensitivity Study A study was conducted to demonstrate that the break location utilized for the preceeding analyses is indeed the worst break location. As stated previously, the analyses were performed assuming that the break was 1,eektad in the bottom of the pump discharge piping. A 0.075 ft 2 bot leg break was analyzed to provide a direct comparison to a similar case in the cold leg. For this evaluation, the RC pumps were assumed to trip after the RCS void fraction reaches 90%. Figure 2.15 shows the average system void fraction transient and the core uncovery times for _
2 both the 0.075 ft hot and cold leg breaks. As shown, the cold leg break l reaches 90% void fraction approximately 150 seconds earlier than the hot leg break. Also, the cold leg break yields a core uncovery time of 175 seconds longer than the hot leg break. The quicker core recovery time for the hot leg break is caused by the greater penetration of the HPI fluid for this break. For a cold leg break in the pump discharge piping, a portion of the HPI fluid is lost directly out the break and 1s,not available for core refill. For a hot leg break, the full HPI flow is available for core refill. Thus, as shown by direct comparison and for the reasons given above, hot leg breaks are less severe than breaks in the pump discharge piping. 8 Peak Cladding Temperature Assessment As described previously, a RC pump trip, at the time the RCS void - , fraction is 90%, will result in core uncovery times of approximately 600 seconds. The peak cladding temperatures for these cases were
, evaluated using the small break evaluation model core power shape used j l to demonstrate compliance with Appendix K and 10CFR50.46. Also, an I
. 1 ediabetic heatup assumption during the time of core uncovery was utilized. l This approach is extremely conservative in that the power shape and
. 1
local power rate (kw/ft) analyzed is not expected to occur during normal plant operation. Furthermore, use of an adiabatic heatup assumption neglects any credit for the steam cooling that will occur during the core refill phase and also neglects the effect of any radiation heat transfer. Using a decay heat power level based on 1.2 ANS at 1500 seconds, the cladding will heatup at a rate will be 6.5 F/S under the adiabatic assumption. With a core uncovery period of j l 600 seconds and the adiabatic heatup assumption, cladding temperatures will exceed the criteria of 10CFR50.46. Use of a more realistic heat - i transfer approach with the extreme power shape utilized for this eval-untion is also expected to result in cladding temperature in excess of the criteria. In order to ensure compliance of the 177 FA lowered loop plants to the criteria of 10CFR50.46 a prompt tripping of the RC pumps is required. Section B. demonstrates that a prompt trip of . the RC pumps upon receipt of a low pressure ESFAS signal vill result l I in compliance to the criteria. . An evaluation of the peak cladding temperature using a power shape encountered during normal operation for a realiktic transient response with delayed RC pump trip will be provided by September 15, 1979. 4 0
- 11'-
C. Analysis Applicability to Davis-Besse I, The significant parametric differences between the raised-loop Davis-Besse I plant and the preceeding generic lowered-loop analysis are in the high pressure injection (HPI) delivery r'ste and the amount
, of liquid volume which can effectively be used to cool the core.
The liquid volume differential is due to the basic design difference; raised versus lowered loops. Because of the raised design, system water available after the RC pumps trip will drain into the reactor vessel. For the lowered loop designs, the available water .s split between the reactor vessel and the pump suction piping. Thus, for the same average system void fraction, the collapsed core liquid level following an RC pump trip is higher for the raised loop design than for the lowered loop design. Figure 2-16 shows a comparison of the delivered HPI flow for the Davis-Besse I plant and the lowered loop plants. As shown, for a similar number of HPI pumps available, the Davis-Besse I pumps will deliver more flow. For the delayed pump trip cases presented in section B.4 of this report, the Davis-Besse I plant will take approximately 450 seconds to recover the core as opposed to :600 seconds for the lowered-loop 7 lants. However, it is noted that the core recovery time is based on us,ing two HPI's rather than one, as required by Appendix K. Use of only one HPI pump for Davis-Besse I will result 1., core uncovery times - in excess of 600 seconds. The Davis-Besse I plant cannot be shown to be l in compliance with 10CFR50.46 for a delayed RC pump trip. ' Prompt reactor coolant pump trip is, therefore, necessary to ensure e compliance of the Davis-Besse I plant with 10CFR50.46.
I
.D. Effect of Prompt RC Pump Trip on Low Pressure ESFAS Signal l l
As demonstrated by the previous sections, the ECC system can not be demonstrated to comply with 10CFR50.46 using present evaluation techniques and Appendix K assumptions under the assulaption of a delayed RC pump trip. Thus, prompt tripping of the RC pumps is necessary to ensure conformance. Operating guidelines ior both LOCA and non-LOCA events have been developed which rgquire prompt i tripping of the RC pumps upon receipt of a low pressure ESFAS signal. Because no diagnosis of the event is required by the operator and ESFAS initiation is alarmac in the control room, prompt tripping of the RC pumps can be assumed. i The effect of a prompt reactor coolant pump trip on se ESFAS signal has been m ained to ensure that the consequences of a small LOCA are bounded by previous small break analyses 2which marrla RC pump trip on reactor trip.- As shown by Table 2-3 at the time of low pressure ESFAS initiation, keeping the RC pumps running results in a Aower average system void fraction. This occurs because the availability of the RC pumps results in lo.eer hot leg temperatures and thus less flashing in the RCS at a given pressure. Thus, a prompt trip upon receipt of an ESTAS signal will result in a less severe system void fraction evolution than cases previously analyzed assuming RC pump on reactor trip. E. Conclusions The results of the analyzes described in this section can be summarized as follows:
- 1) If the RC pumps r== min operative, core cooling is assured regardless of system void fraction.
2
- 2) For breaks greater than 0.025 ft , the RCS may evolve to system void fractions in excess of 90%.
- 13 . .
. l 2
- 3) At 40 minutes, the 0.025 f t break has evolved to only a 47% void 2
fraction. Thus, a delayed RC pump trip for breaks less than 0.025 f t will not result in core uncovery.
- 4) The potential for high cladding temperatures for a small break transient with delayed RC pump trip is restricted to a time period a
between that time where the system has evolved to a high void fraction and the time of LPI actuation.
- 5) Even with 2 HPI pumps available, tripping of the RC pumps at the worst time (90% void fraction) results in a core uncovery period which cannot be shown to comply with 10CFR50.46, if Appendix K _
assumptions are utilized.
- 6) A prompt RC pump trip upon receipt of a low pressure ESFAS signal ,
will provide compliance to 10CFR50.46.
- 7) The above conclusions are applicable to both the B&W 177 FA lowered and raised loop NSS designs.
l
. 1 O
A 1 l 8 . l
i i 4 III. IMPACT ASSESSMCTT OF A RC PUMP TRIP ON NON-LOCA EVDrTS ! , f' *
\
I A. Introduction I
- Some Chapter 15 events are characterized by,a primary system I
i response similar to the one following a LOCA. The Section 15.1 events that result in an increase in heat removal by the secondary systes cause a primary system cooldown and depressurization, much like a small break LOCA. Therefore, an assessment of the conse-quences of an imposed RC pump trip, upon initiation of the low RC pressure ESFAS, was made for these events. B. General Assesseent of Pump Trip in Non-LOCA Events Several concerns have been raised with regard to the effect that an early pump trip would have on non-LOCA events that exhibit LOCA ) characteristics. Plant recovery would be more difficult, dependence. on natural circulation mode while achieving cold shutdown would be highlighted, manual fill of the steam generators would be required, and so on. However, all of these drawbacks can be accouanodated since ( none of them will on its own lead to unacceptable consequences. Also, restart of the pumps is not precluded for plant control and cooldown
'once controlled operator action is assumed. Out of this search,
~
j three major concerns have surfaced which have appeared to be sub-stantial enough as to require analysis:
- 1. A pump trip could reduce the time to system fill /repressurization or safety valve opening following an overcooling transient. If the time available to the operator for controlling HPI flow and the margin of subcooling were substantially reduced by the pump trip to where timely and effective operator action could be questionable, the pump trip would become unacceptable. -
- 2. In the event of a large steam line break (maximum overcooling), the blowdown may induce a steam bubble in the RCS which could impair natural circulation, with severe consequences on the core, es-
, pecially if any degree of return to power is experienced.
- 3. A more general concern exists with a large steam line break at EOL
, conditions and whether or not a return to power is experienced following the RC pump trip. If a return to critical is experienced, natural circulation flow may not be sufficient to remove heat and
- to avoid core damage.
- 15.- ,
l . _ _ _ . -- .
Overheating events were not considered in the impact of the RC pump trip since they dc not initiate the low RC pressure ESFAS, and clierefore, there would be no coincident pump crip. In addi-tion, these events typically do not result in an empty pressurizer
, or the formation of a steam bubble in the prima,ry system. Reactivit-transients were also not considered for the same ieasons. In addi-
. tion, for overpressurization, previous analyses have shown that for the worst case conditions, an RC pump trip will mitigate the pressu:e rise. This results from the greater than 100 psi reduction in pressure at the RC pump exit which occurs af ter trip. C. Analysis of Concerns and Results
- 1. System Repressurization i In order to resolve this concern, an analysis was performed ]
for a 177 FA plant using a MINITRAP model based on the case I set up for IMI22, Figure 3.1 shows the ncding/ flow path scheme used and Tabl,e3,1 provides s description of the nodes' and flow paths. This case assumed that, as the result of a small steam line break (0.6 ft. split) or of some combination of secondary side valve failure, secondary side heat demand was increased from 100% to 138% at time zero. This increase in secondary side heat demand is the smallest' which results in a (high flux) reactor trip and is very similar to the worst moderate frequency overcooling event, a failure of the steam pressure regulator. In the analysis, it was assumed ; that following HPI actuation on low RC pressure ESFA3, main ) feedwater is ramped down, MSIV's shut, and the auxiliary feedwater initiated with a 40-second delay. This action was i taken to stop the cooldown and the depressurization of the I system as soon as possible af ter HPI actuation, in order to , minimize the time of refill and repressurization of the system. Both HPI pumps were assum&d to function. The calculation was perlarmed twice, once assuming two of the four RC pumps running (one loop), and once assuming RC pump trip right after HPI initiation. The analysis shows that the e system behaves very similarly with and without pumps. In both cases, the pressurizer refills in about 14 to 16 minutes from initiation of the transients, with the natural circula- l
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