ML19208A060

Responds to IE Bulletin 79-05C Re RCS Pump Trip & Guidelines for LOCA & non-LOCA Transients Accounting for RCS Pump Trip Requirements.Forwards B&W Rept, Analysis Summary in Support of Early Reactor Coolant Pump Trip

ML19208A060
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/29/1979
From:	Trimble D ARKANSAS POWER & LIGHT CO.
To:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
1-089-16, 1-89-16, NUDOCS 7909120477
Download: ML19208A060 (4)

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1-089-16 Mr.

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Reid h-August 29, 1979 Attachment Oc:

Mr. Victor Stello, Jr.,

Director Office of Inspection and Inforce=ent U.

S.

Nuclear Regulatory Co==.

Washington, D.

O.

20555 Mr.

W.

D.

Johnson U.

S.

Nuclear Regulatory Co==.

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3cx 2090

'Ir. Harold 3.

Denton, Director Nuclear Reactor Regulation U.

S.

Nuclear Regulatcry Co==ission Washington, D.

C.

20555 900RDMWA.

94000 <,

ATTAGNENT 1 ANALYSIS SIM.ARY IN SUFFORT OF AN EAP1Y RC PUMP TRIP ep 9

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i CONTENTS Page I.

INTRODUCTION.

1 II.

SMAI.L 3REAK ANALYSIS.

2 A.

Introduction.

2 B.

System Response With RC Pumps Running 2

C.

Analysis Applicability to Davis-Besse 1 11 D.

Effect of Frenpt RC Punp Trip on Lcw ?ressure ESFAS Signal..

13 E.

Conclusions 13 III.

IMPACT ASSESS INT OF A RC PUMP TRIP CN NCN-LOCA EVENTS.

15 A.

Introduction.

15 3.

General Assessnent of Pump Trip in Non-LOCA E tents.

13 C.

Analysis of Ccncerns and Results.

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16 D.

Conclusions and Summary 18 D " #7' D

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D ANALYSIS SDO!ARY IN SUPPORT OF g bl q

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AN EARLY RC PO!? TRIP d(0d l

us I.

INTRODUCTION 3&W has evaluated the effect of a delayed RC pu=p trip during the course of s=all loss-of-ecolant accidents and has fcund that an early trip of the RC pumps is required to shew confor=ance to 10CFR50.46, A su==ary of the LOCA analyses perfor=ed to date is provided in Section II.

This discussien includes :

1.

A description of the models utilized.

2.

Break spectru= results wi:h continuous RC Pu=p Operation.

3.

Break spectru= results with delayed RC pu=p trips including estinates of peak cladding te=peratures.

4.

Justification that a prc=pt pu=p trip follcwing ISTAS actuatien on icw RC pressure provides LCCA =itigation.

An impact assessment cf :he required pu=p ::1p en non-LCCA events has also been cc=pleted and is presented in Secticn III. This evaluatica supper:s the use of a pu=p : rip following ISFAS actuaticn for LOCA Nitigaticn since no detri= ental consequences en non-LOCA events were identified. 3 % 060

II.

SMAI.L 3REAK ANALYSES A.

Introduction Previous small break analyses have been perfor=ed assu=ing a less-of-offsite power (reactor coolant pump coastdown) coincident with re-actor trip.

These analyses support the conclusion that an early RC pump trip for a LCCA is a safe condition. However, a concern has been identified regarding the consequences of a s=all break transient in which the RC pu=ps remain operative for some ti=e period and then are lost by so=e =eans (operator action, less-of-offsite power, equip =ent failure, etc.).

This section contains the results of a study to further understand how the small break LOCA transient evolves with the RC pu=ps operative.

Specifically, section 3.

describes the syste= respense with the RC pumps running for B&W's 177-FA lowered-loop plants.

In-cluded in this section is the develop =ent of the codel 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 de=enstrates the applicability of the conclusions drawn in section ' 3 to a 177-FA raised-loop plant (Davis-3 esse 1).

The effect of a prc=pt tripping of the RC pucps upon recei; of a low pressure ESFAS signal is discussed in section D Finall2, se:-

ion E. su==arices the conclusicas of this analysis.

3.

Systen Rescense With RC Puces Runnine 1.

Introduction Recent evaluations have been performed to examine the primary syste= response during small breaks with the RC pu=ps opera:ive.

During the transient wi:h the RC pu=ps available, the forced circulatien of reac cr coolan: vill maintain the core at or near I

em em g

D the saturated fluid temperature.

However, for a range of break es y; sices, the reactor ecolant systen (ICS) cil evolve to high void gg PD r "h} ' 1[

A fractions ue : the slew systen depressurication and the high JQ]

- (D JT

.;] liquid (low quali:y fluid) cischarge through the break as a re-suit of the forced circulation.

In fact, the RCS void fraction will increase :o a value in excess of 9C% in the short ter=.

In e

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4

the 1cng ters, the system void fraction will decrease as the RCS RCS depressurizes, HPI flew increases, and decay heat diminishes.

With the RCS at a high void fraction, if all RC pu=ps are postu-lated to trip, the forced circulation will no longer be available and the residual liquid would not be sufficient to keep the core' covered. A cladding temperature excursion would ensue until core cooling is reestablished by the ECC systems. The following para-graphs su==arizes the results of the analyses which were perfor=ed for the 177-FA lowered-leep plants, to develop the consequences of this transient.

2.

Method of Analvsis The analysis =ethod used for this evaluation is basically that de-scribed in section 5 of 3AW-10104 Rev. 3, "3&W's ECCS Evaluatica Model"1 and the letter J.H. Taylor (35%) to S.A. Varga (:!RC), da:ed 2

..uly 18, 1978, which is applicable to the 177-FA lowered-loop plants for pcwer levels up to 2772 rit.

The analysis uses the CRAFT 23 code to develop the history of the RCS hydrodyna ics.

However, the CRAFT 2 model used for this study is a modifica: ion of the small break eval,uation nedel described in the abcve ref-erences.

Figure 2-1 shews the CRAFT 2 neding diagram for small breaks from the above referenced letter.

The nodified CRAFT 2 model consists of 4 nodes to si=ulate the pri=ary side, 1 nede for the secondary side of the steam generator, and 1 node. representing the reactor building.

Figure 2-2 shews a sche =atic diagras cf this =cdel.

Node 1 contains the ccid leg pt p discharge piping, downconer, and lower plenu=.

cde 2 is the pri=ary side of the SG and the pu=p sucticn piping.

',; ode 3 contains the core, upper ple-num, and the het legs.

Nede 4 is the pressuri:er and nodes 5 and -

6 represent the reac:c: building and :he SG secondary side, re-spectively.

This 6 nede ecdel is highly simplified ccupared :c those utilized in pas ICCS analyses.

does, however, main: sin RCS volume and elevatica rela:ionships which are imper: ant o properly evaluate the syste= respcase during a small break wi:h the RC pumps running.

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9 The breaks analy:ed in this section are assumed to be located in the told leg piping between the reactor coolant pump discharge and the reactor vessel.

Section 3.7 de=custrates that this is the worst break locatice Kef assu=pticas 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.

Ecuinment Availability The analyses which were performed assu=ed 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 pu=ps were assumed to trip. Also, the i=-

pact of 1 versus 2 E?I syste=s for pu=p injection were exa=ined.

The majority or the analyses perfer ed assumed 2 E?! pumps.

However, as is demonstrated later, even with 2 EPI pu=ps avail -

able, cladding temperatures will exceed the criteria of 10 CFR 50.46 using Appendix K evaluatica techniques.

Therefore, fur-ther analysis with enly 1 EPI pu=p would only be academic.

b.

Phase Se:aration The presen: ECCS evaluatien =cdel created to evaluate small breaks without RC pu=ps operative,(quiescent RCS) uti-li:es the Wilson" bubble rise correlation for all primary sys-tem contrcl volumes in the CRAFT evalua:icn. In this analysis, g3 g3 for the time period that the RC pumps are operative, the pri-n

=ary system ecolant is assumed to be hecc3eneous, i.e.,

no CJ CD E

- phase separation in the systa=.

In reality, the flew rates Cd _3) '_ #h-)

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in the core and act legs are le' enough that s. lip will eccur.

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vessel co= pared to that calculated with the homogeneous :cdel.

With the he:cgenecus assu=ption, core fluid is centinucusly circula:ed throughout the pri:ary syste: and a portien of that fluid is los: via :he break.

During the later stages of the transient, a slip =cdel will resul: in fluid being trapped in the reactor vessel and the ho: legs.

The only methed of icsing liquid during :his period will be by boiling caused by the core decay heat.

Thus, the assumption of he:cgeniety for the peried with the RC pu=ps operative is ccnservative.

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Following tripping of the RC pu=ps and the subsequent loss-of-forced circulation, the syste: 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 ho=ogenecus assu=ption for the period with the RC pu=ps operating applies to nodes 1, 2, and 3 in the CRAFT =cdel.

Node 4, the pressurizer, and node 6, the secondary side of the stea: generators, utill:e the Wilson bubble rise codel throughout the transient as these nodes are not in the direct path of the forced circulation.

3.

Bench =arkine of the 6 Node CRAFT ::odel Studies uere perfor:ed Oc compare the results of the 6 nede codel to the core extensive evaluation ecdel fo: 3&U's 177-FA lowered-loop plants as described in the le::er J.H. Taylor (B&U) to S.A.

Varga (NRC), dated July IS,1978.

The break size selected fer 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 vieued as " realistic" or the wors: that would be ex-pected en a real plant.

Fig wes 2-3 and 2-4 are the results of

~{ Ch Cd D

this comparisen.

Systen pressure and percent void fraction shewn e, j[

in Figures 2-3 and 2-4, respectively, compare very well wi:h those q)

F}

from the scre extensive (23 nodes) CRA:C2 s all break acdel. As g1 F

-(DJ a )(; _k

.lgen in these figures, the difference is not significant and is 3

,us less than a few percent.

The computer :ine for this 6 nede scdel is, however, significantly decreased.

The cdel utili:ed for this study is thus justified based en comparisen of resul:s to the core ex:ensive small break model and desirable because of its ecencrical run :ime.

4 Analvsis Results The break sizes examined for :his analysis ranged frc 0.025 f:2 to 0.2 f:, in area and are located in the pump discharge piping.

3reaks of this size do no: result in a rapid system depressuri-

=ation and rely predc=inantly upon :he E?2s for =itigation.

5-9Ar II) '

=

Table 2-1 su==ari:es the analyses performed for this evaluatien.

The majority of the analyses performed utili:ed 2 HPI pumps through-out the transient.

The effect of utilizing i HPI pu=p is discussed in this section.

Figures 2-5 and 2-6 shou the system pressure and average system,

void fraction transients for the break spec: rum analyzed assuming continuous RC pump operation and 2 HP!'s available.

In Figure 2-6, the average system void fraction is defined as V

-V 1

2 Average system void, ; =

x 100 yI where v'1 = total primary liquid volume excluding the pressuri-

er at time = 0, V, = total primary liquid volume excluding the pressuri-

er at time =

This parameter was utili:ed in place of the mixture height in tha:

the coolant will tend to be he=ogencousif =ixed with the RC pu=ps operative. Under these assumptions, the core is cooled by forced q

circulation of two-phase fluid and not by pool boiling,as in :he p

case where the RC pu=ps are not running and separatien of stes:

(

'3 and water occurs. As shown in Figure 2-3, the syste: pressure re-

/i i

t z spense is basically independent of break si:e during the first U

3 several hundred seconds into the transient.

This occurs because h\\ '

Q the forced circulation of reactor coolant maintains adequate heat g

transfer in the stea: generators; the primary sys:en :hus depres-suri:es :o a pressure (about 1100 psia) corresponding :o :he sec-ondary control pressure (i.e., set pressure of SG safety relief valves). After some time (250 seconds for the 0.1 f:2 break), the system pressure will decrease as the break alone relieves the core energy.

Figure 2-6 shows :he evolution of the system void fractien; values in excess of 902 are predic ed very earl / (300 seconds) into the transient.

For the larger breaks the system high void fractions occur early in time.

yor the smaller breaks it takes in the order of hours before the system evolves :o high void fraction.

Core cooling is naintained during a small break with continuous RC pump 3d6066 6m he.

operation regardless of void fraction.

In the long ter=, the sys-te will depressuri:e and the enhanced performance of the ECCS (HFI and LPI) will result in reduced syste= void fraction.

Figure 2-7 illustrates this long ter syste= behavior for a 0.10 ft2 t /'

break.

For this case, the LPIS are operative at approxi=ately 2300 ?.

seconds, anc a substantial decrease in syste= void fraction results.

An arbitrary pu=p trip after approxi=ately 2700 seconds would not result in core uncovery. The potential for core uncovery due to

.Y an RC pump trip i.; thus li=ited to a discrete time period during V

N g

which the natural evolutien of the syste= prede.es high void frac-h tiens and prior to LPI actuation.

For a 0.1 f:- break, this ti=e

\\ ^J 4 ; h>/

s period is en the order of 2000. seconds.

For s= aller breaks, this critical time could be a few hours even if the operator initiatec

?,, a y

7 a centrolled cooldown and syste= depressurization as recc== ended

/

in the small break guidelines.

.r Although the analyses described above used 2 HPI pu=ps, the effect of only 1 HPI pu=p available on the syste= void fractica evolutien while the RC pu=ps are operating is not significant.

Figures 2-3 and 2-9 shew the i= pact of one versus vo HPI pumps on syste= pres-sure and avera;e void fr:c:icn transients for a 0.05 ft2 break vi h the RC pumps operative. As seen frc= these figures, the results with one HPI pu=p are not significantly different :o the two HPI pu=p case and are bounded by the spectru= approach utilized. With one HPI pu=p,. the syste= does depressurize = ore slowly (less stea:

condensation) and a higher shcr: ter ecuilibriu voi! fraction a

is achieved. Also, reccvery of :he core following a less of the m

Q RC pu=ps would be significantly longer with only 1 HPI pu=p avail-able.

M ca

%]

The majority of the analyses provided in this report uses two H?!

pu=ps and denenstrates a ccre c clin; pr ble= with ecrs: ti=e pu=p trip given : hat assu=p:icn.

As analysis of One 3?! available cases would only show a larger proble=, such cases have not been exten-sively considered.

As de=onstrated in section 3.4, the resolutien of this proble=, forced early pu=p trip, provides assurance of core cooling for both cne or two IPIs available cases. Therefore, M6066

there is no need for further pursuit of the single EPI available case.

The effect cf the RCP tripping during the transient was studied by assuming that the pu=ps are lost when the systes reaches 90% void fraction. Loss of the RC pumps at this void fraction is expec:sd to produce essentially the highest peak cladding ta=perature.

After the RC pu=ps are tripped, the fluid in the RCS separates and liquid falls to the lowest regions, i.e.,

the lower plenum of the RV and the pu=p suction piping. At 90% void fraction, the core will be totally uncovered following the RC pu=p trip.

Thus, the time required to recover the core is longer than that for RC pu=p trips initiated at Icuer system void fractions.

System void frac-tions in excess of 90% can possibly result in slightly higher tes-peratures due to the longer core refill times that =ay occur.

However, the peak cladding temperature results are not expected to be significantly different as the syste= 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 with the RC pu=ps tripping at 90% void fraction with 2 EPI pumps avail-able for core reccvery. As shewn, the core will be uncovered for approximately 600 seconds for the breaks analy:ed.

Figures 2-10 and L-F"'show the system pressure and void fraction response for 3

(h*'h the'0 0 5 f *b break with a RC pu=p trip at 90% void fraccion. As c

. 7-'

seen...hese figures, the systes depressuri:es faster after the v

/

[/y RC pu=p trip, due to the change in leak quality, and the void 1.

s.

fraction decreases indicating that the core is being refilled.

,#J>

Figure 2-12 shows :he core liquid level response following the RC

( g%

Y

(

.p*

q pu=p trip.

The core is refilled :o the 9 foot level with collapsed u-

+

e t*

liquid approxi=ately 625 seconds after :he assu=ed pu=p trip.

t 2

1 Cnce the core liquid level reaches the 9 foot elevatien, the core

\\

L\\

i s

f is expected to be covered by a tuc-phase =ix:ure and the cladding

(( g te=perature axcursion would be ter=ina:ed.

4 F

D D

=

.' N 3 s<s J

bD m

~

~

A s

O A l

_, J LI

-7

.a-2 067

5.

Effect of 1.0 ANS versus 1.2 ANS Decav 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%

system void fraction.

Figures 2-13 and 2-14 show a comparisen of system pressure 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 dr:p from caturation pressure (s1100 psia) about 200 seconds earlier than the case with 1.2 /3S as a result of reduced decay heat.

Also, the system will evolve to a lower average void fraction as snown ia Figure 2-14 After the pumps trip at 90% system void frac:fon, the case with 1.0 ASS decay curve has a shorte.: core uncovery time by approxi=stely 200 sec-onds co= pared to 1.2 ANS case.

This case demonstrates that the effect of a delayed RC pump trip may be acceptable when viewed realistically. A peak cladding temperature assessment for this case will be provided in a supplementary response planned for September 15th, to the IiE Eulle:in 7905-C.

6.

Effect of No Auxiliarr Feedwater Analyses have also been performed with the RC pumps available and no auxiliary feedwater. These analyses all assumed 2 HPI pumps were available.

The syste: void frac.on evolutions for these calculations were not significantly different fro: those discussed with auxiliary feedwater.

Thus the conclusions of the cases with auxiliary feedwater apply.

C3C31 0

es es

~

S,

-.3

.t

- 9 % 968

.2 Break Location Sensitivity Study A study was conducted to de=onstrate that the break location utilized for the preceeding analyses is indeed the worst break locatien.

As stated previously, the analyses were performed assuming that the break 2

was icgated in the botto= of the pu=p discharge piping. A 0.075 ft hot leg break was analyzed to provide a direct co=parison to a similar case in the cold leg. For this evaluation, :he RC pu=ps were assu=ed to trip after the RCS void fraction reaches 90*..

Figure 2.15 shows the average syste= void fracticn transient and the core unccvery ti=es for both the 0.075 ft hot and cold leg breaks. As shown, the cold leg break reaches 90*: void fraction approximately 150 seconds earlier than the hot leg break. Also, the cold leg break yields a core uncovery ti=e of 175 seconds longer than the hot leg break.

The quicker core recovery

~

time for the hot leg break is caused by the greatar penetration of the w

%Q HPI fluid for this break.

For a cold leg break in the pu=p discharge

(

)i piping, c pertion of the HPI fluid is los: directly out the break and i

b b

is not available for core refill.

For a hot leg break, the full H?I i

i SM fluw is available for core refill.

Thus, as shown by direct co=parisen (d

and for the reasons given above, hot leg breaks are less severe :han breaks in the pu=p discharge piping.

8 Peak Cladding Te=rerature Assess =ent As described previously, a RC pu=p trip, at the ti=e the RCS void -

fraction is 90*, will result in core uncevery :i=es of apprcxi=ately 600 seconds.

The peak cladding te=peratures for these cases were evaluated using the s=all break evaluatice =cdel core power shape used to de=enstrate cc=pliance with Appendix K and ICCFR50.4c.

Also, an adiabatic heatup assu=ption during the ti=e of cere uncevery was utill:ed.

This apprcach is extre=ely conservative in that the pcwer shape and

. 3 % 069

local power rate (kw/ft) analyzed is not expected to occur during normal plant operation.

Further= ore, use of an adiabatic heatup assu=ption neglects any credit for the stea: 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 s rate will be 6.5 F/S under the adiabatic assu=pticn. With a core uncovery period of 600 seconds and the adiabatic heatup assu=ption, cladding temperatures will exceed the criteria of 10CFR50.46.

Use of a more realistic heat transfer approach with the extre=e power shape utilized for this eval-uation is also expected to result in cladding te=pprature in excess of the criteria.

In order to ensure cc=pliance of the 177 FA lowered loop plants to the criteria of 10CFR50.46 a prc=pt tr y " ; i the RC pu=ps is required.

Se::icn 3. de=ctstrates ~

rip of a/

the RC pu=ps upon receipt of a Icw pressure E FAS dignal will result in ec=pliance to the criteria.

An evaluation of the peak cladding te=perature using a power shape encountered curing nor=al operation for c re/

alistic transient response with delayed RC pump trip will be pr:: vide ' by Septe=ber 13, 1979.

.sC D

D IC e

c QQ e-m

\\,.

O' 0

A j

'v a

1. A a O'16Ui'O

C.

Analvsis Acolicabi? -* to Davis-3 esse I The significant parametric differences between the raised-loop Davis-Sesse I plant and the preceeding generic lowered-lcop analysis are in the high pressure injection (HFI) delivery rate and the a: cunt ofliquidvolumdwhichcaneffectivelybeusedtococlthecobe.

The liquid volu=e differential is due to ne basic design diffarence; raised versus lowered loops.

Because of the raised design, system water available af ter the RC pumps trip will drain into the reactor vessel.

For the lowered loop designs, the available water is spli:

between the reactor vessel and the pu:p suction piping.

Thus, for the same average system void fractica, the collapsed core liquid level following an RC pu=p trip is higher for the raised icop design than for the lowered loop design.

~

Figure 2-16 shcws a compariscn of the deli'rered HPI flew for the Davis-3 esse I plant and the 1:wered locp plants.

As shown, for a sd-dia nu:ber of EFI pumps available, the Davis-3 esse I pumps will deliver more flow.

Fcr the delayed pu=p trip cases presented in sec:icn 3.4 of this repor:, the Davis-Besse I plant will :ake approximately 450 9

seconds to recover the core as opposed te :600 seconds for the lowered-i

-c g3g3 loop plants.

Ecwever, it is noted :ha: the core recovery :::e is based

{ }, M en using two EPI's rather than ene, as required by Appendix K.

Use of h

)(21Cd only one EPI pump for Davis-Besse I will result in core uncovery-times r-i C

{g;;;n]

in excess of 600 seconds.

The Davis-3 esse I plant cannot be shewn to be

(

]

in compliance with ICCFRf0.46 for a delayed RC pump : rip.

Pro:p reactor coolant pump trip is, therefore, necessary to ensure cc=pliance of the Davis-Bessc : plan with ICCFRf0.46. f'

• O 'Sh) f y{* v,9 U

.D.

Ef f ect of Protet RC Puto Trio on Lew Pressure ESFAS Sienal As deconstrated by the previous sections, the ECC syste: can not be de=cnstrated to comply with ICCFR50.46 using present evaluation techniques and Appendix K assumptions under the assumption of a delayed RC pump trip.

Thus, prc=pt tripping of the RC pumps is necessary to ensure conformance.

Operating guidelines for both LOCA and non-LCCA events have been developed which require prc=pt tripping of the RC pumps upon receipt of a low pressure ESFAS signal.

Because no diagnosis of the event is required by the cperator and ESFAS initiation is alar =ed in :he control rec =, prompt tripping of the RC pumps can be assumed.

The effect of a pronp: reac:or ecclant pump trip on an ESFAS signal has been examined to ensure that the consequences of a stall LCCA are beunded by previous small break analyses which Issume RC pump trip en reactor trip.

As shown by Table 2-3 at the ti=e of icw pressure ISFAS initiation, keeping the RC pumps running results in a lower average syste: void fraction.

This occurs because the availability of the RC

,a pr,ef pumps results in lower hot leg te:peratures and thus less flashing in the RCS at a given pressure. Thus, a prompt trip upon receipt of an G - ) @7:

e ESFAS signal will resul: in a less severe syste void fractica evolution i

a W

io than cases previcusly analyzed assuming RC pu=p on reactor trip.

)

E.

Co nc lu s ien_s, The resui:s of :he analyzes described in this sectica can be sunnariced as folicws:

1)

If the RC pu:ps remain operative, core cooling is assured regardless of systen veid fraction.

n 2)

For breaks greater than 0.025 f:~, the RCS =ay evolve :c syste: void fractions in excess of 9C*.

o.

1. y.

....,o

!Xt O. (,%r

~~

9

3) At 40 minutes, the 0.025 f t' break has evolved to only a 47% void n

fraction. Thus, a delayed RC punp trip for breaks less than 0.025 ft' will not result in core uncovery.

4) The potential for high cladding te=peratures for a small break transient with delayed RC pu=p trip is restricted to a time p riod between that ti=e where the systa= has evolved to a high void fraction and the time of LPI actuation.

5) Even with 2 RPI punps available, tripping of the RC pu=ps at :he wors: time (90% void frac:lon) results in a core uncovery period which cannot be shown to co= ply with ICCFR50.46 if Appendix K assumptions are utili:ed.

6) A pro:pt RC pump trip upon receipt of a low pressure ISFAS signal will provide compliance to 10CFR50.46.

7) The above conclusions are applicable :o both the 36W 177 FA lowered and raised loop NSS designs.

r3 r3 9

9

<> av g

.,D' sj

_ [

0 o

__3 s4co7a

III. IMPACT ASSESSMENT OF A RC PUMP TRIP ON NON-LOCA EVENTS A.

Introduction Some Chapter 15 events are charac:cri:ed by a primary system response similar to the one following a LOCA. The Section 15.1 events that result in an increase in heat removal by the secondary system cause a pri=ary system cooldown and depressuri:stion, much like a small break LCCA. Therefore, an assessment of the conse-quences of an imposed RC pu=p : rip, upon initiation of the low RC pressure ESFAS, was made for these events.

3.

General 3.ssessment of Pumo Trio in Non-LCCA Events Several concerns have been raised with regard to the effect : hat an early pu=p trip would have en non-LCCA events that exhibit LCCA characecristics.

Plant recovery would be cre difficult, dependence.

on natural circula:ica code while achieving cold shutdcun wculd be highlighted, manual fill of the stea= genera:crs would be required, and so on.

However, all of these drawbaci:s can be accc :cda:cd since none of the:.till en its own lead to unacceptable ccusacuences.

Alsc, restart of the pu=ps is not precluded for plant control and cooldewn once controlled cperater action is assumed. Cut of this search, three ajor concerns have surfaced which have appeared to be sub-i-

9 stantial enougb as to require analysis:

-a 1.

A pu=p trip could reduce :he time to systen fill /repressuri:atien pg1;J G

9,b[2f or safety valve cpening fellcwing an overcooling :ransient.

If the rise available :o tha operator for con:rciling RFI flew and D) h3' the acrgin of subccoling were substantially reduced by the pu=p ca

{ggg;]

trip to where timely and effective operator actica ceuld be

--q questionable, the pu=p trip wculd becc e unacceptable.,

2.

In the event of a large s:ea= line break (naximu: overcooling), :b blowdown ay induce a steam bubble in tha RCS which could impair natural circulatien, with severe consequences en the cere, es-pecially if any degree of return :o pcwer is experienced.

3.

A =cre general concern exists wi:h a large steam line break at ICl conditions and whether or not a return to power is experienced folicwing the RC pu=p trip.

If a return to critical is experienec natural circulation ficw may not be sufficient to re=cve heat and to avoid core damage.

34.607.1

-.o -

Ovc henting events were not considered in the i= pact of the FC pump trip since they do not initiate the low RC pressure ESFAS, and therefore, there would be no coincident pump trip.

In addi-tion, these events typically do not result in an empty pressurizer or che formation of a stea= bubble in the pri=ary syste=.

Reactivity transients were also not considered for the same reasons.

In addi-tion, for overpressurization, previces analyses have shewd that for the worst case conditions, an RC pu=p crip will mitigate the pressure rise. This results frem the greater than 100 psi reduction in pressure at the RC pump exit which occurs after trip.

C.

Analvsis of Ccncerns and Results 1.

System Recressuri stien In order to resolve this conectn, an analysis was perfor=ed for a 177 FA plant using a MI"ITRA? =edel based en the case set up for T::I-2.

Figure 3.1 shows the neding/ficw path sche =e used and Tabl,e 3.1 prevides s descriptien cf the nodes' and flew paths. This case assumed tha:, as :he result of a s=all stea line break (0. 6 f t.

spli:) or of se=e cc binatic

~

of secondary side valve f ailure, secondary side heat d emand was increased frc= 100% :o 138% at ti=e zero.

This increase t:7g]

in secondary side heat demand is the smalles: which results in a (high flux) reac:c: trip and is very s1=ilar to :he

-c ca worst moderate frecuency evercoolin; event, a failure cf the cra C 2:3 steam pressure regula:cr.

In the analysis, it was ass==ed h2IE3 that follouing H?I actuation en icw RC pressure ISFAS, main C)

--.)

ca C:3 f eedwater is rc ped deur., ' MSIV's shut, and the auxiliarv.

lC feedwater initiated wi:h a 'C-second delay. This actica was C) taken to step the cocidewn and the depressurizaticn of the system as soon as possible af:er HPI actuation, in crder to mini =inc the cine of refill and repressurizatica cf the systen.

Both EP pumps were ass = 4d :o functien.

The calculation was perfen:ed twice, once assuming :wo cf :P four RC pumps running (One loop), and once assuming 30 purp trip right af cr EPI initiation.

The analysis shcus that th systen behaves very s1=ilarly with and without pu=ps.

In both cases, the pressuri:cr refills in about la Oc 16 r.inu:c frc= initiation of the transients, with the natural circula-, rp fr c I<LG L dt) e

tion case refilling about one minute before the case with two of fcur pumps running (See Figures'3.2,3.3). In both cases, the systen is highly subccoled, frc= a mini =um of 30*? :o 120 F and increasing at the end of 14 =inutes (ref er to Figure 3.4).

It is concluded that an RC pu=p trip following HPI actuation will not increase the probability cf causing a LOCA through the pressurizer code safeties, and : hat the operator will have the same lead ti=e, as well as a large margin of subecoling, te control HPI prict to saf ety valve tapping. Although ne case with all RC pu=ps was nade, it can be inf erred f rem the one loop case (with pu=ps running) that the subccoled cargin will be slightly larger for the all pu=ps running case.

The pressuri:er will take longer to fill but should de so by 16 minutes into the transient.

Figure 3.t shews the coolant temperatures (hot leg, cold leg, and core) as a function of et:e for the _no RC pumps case.

2.

Effect of Stear Subble en Natural Circulatica Cceline For this cencern, an analysis was perfer:ed for the sa:e n

generic 177 FA plant as cu: lined in Part 1, but asscning that

'l i

as a result of an unmitigated large S13 (12.2 f:.' OER), the ca c:q excessive cooldewn would produce void f ermatien in the primary M

{

g systc=.

The intent cf the analysis was to also show the extent of the void f ert::icn and where it occurred.

As in b

9 bg}43,<

the case analyzed in Par: 1, the break was sy==e ric to bcth Fa -)

r fc,a generators such that both would blew dcwn equally, maxi =1:ing

___, )

the cocidewn (in this case there was a 6.1 f:.2 break en each leop).

There was nc ':SIV closure during :he trnnsient en either steam generater : =aximize cooldewn.

Alsc., the tur-bine bypass system was assured Oc operate, upon rupture, until isclation en ESFAS.

ISFAS was initiated en icw RC pressure and aise actuated H?' (beth pu=ps), tripped RC wnen appa.,cac_,c) ana 1sc,atc;. One

>.24 <v_......s.

..n e.s. -..

e pumps s was ini:ia cd to be:S genera:crs en the icw SG pressure signal, with minimus delay time (both pumps cperating).

This analysis was perforned twice, once assuming all RC pu=ps running, ence with all pumps being tripped en the HPI actuatien (af ter ESFAS), with a sher

(^.5 second) delay.

In both cases, voids were formed in the het legs, but the dura-

$d'o-(_y~s 1)

-t'-

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 cecidown rate, there was less void for=a-tien, resulting f rem the additional system mixing.

The coolant tenperatures in the pressurizer loop hot and cold legs, and the core, are shown for both cases in Figures 3.5 3.6.

The core outle: pressure and SG and pressurizer levels versus ti=e are given f or both cases in Figures 3.S, 3.9.

This analysis shows that the system behaves very si=ilarly with and without pumps, although =aintaining RC punp flow does seem to help =f.tigate void formation. Th pump flow case shows a shorter ti e to the start of pres-surizer refill than the natural circulatien case (Figure 3.

although the :L:e dif f erence does not seem to be very large 3.

Effect of Return to pouer There was no return to pcuer exhibited by any of the ECL cases analyzed above.

Previcus analysis experience (ref.

Midland FSAR, Sectica 132) has shown that a RC pump trip wt

=itigate the censequences of an-ECL return :o power condi::

]

by reducing the cecidewn of the primary system.

The reduct cooldown substantially increases the suberi:ical cargin

_c

[h5,, which, in turn, reduces or eliminates return :o pewer. { ], D. Conclusions and Sennarv 3 {, j(2[$} A general assessment cf Chapter 15 non-LCCA events identi. L_- J three areas that warranted further investiga:icn f or i= pact of C3 (hE. a RC pump trip en ISFAS low RC pressure signal. 1. It was found that a pump trip does net significantly shcr: the time to filling of :he pressurizer and approxi ately :: same ti=e interval fer operator actica exists. 2. For the maximum overeccling case analyzed, the RC pump tri: increased the a cun: cf twc-phase in the primary loop; however, the percent icid fermatice is still :co s=all :o affect the abili:y to ceci en natural circula:1cn. 3. The suberiticci return-to-pcwer condition is alleviated by the RC pump trip case due to the reduced overcooling effee Based upcn :he abcvc assessment and analysis, it is cen-cluded tha: the censecuences of Chapter 13 non-LOCA events are ..g .j $ h dr,_/ # ~ " ~

increased due to the addition of a RC pump trip en 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 analyses for the lowered loop plants are applicable. A. N G 9u R @Ej f a, u'_ z' b_D b 9 9 9 e 9%O78 _ 19 -

Table 2-1. Analysis Scoce With A W Available Continuous RC L*" E#E #E Break size, (ftl) Cold In Hot leg 2 HPI 1 HPI 2 HPI 0.025 X X O.05 X X* X X* 0.075 X X X X 0.10 X X X 0.20 X X

• Analy:ed with both 1.0 and 1.2 ANS decay curves.

DO u0O D L O Dl n bu[t)l,GT J b]1 4 3M;O79 Table 2-2. I= pact Assess =ent of Break Spectrum With RC Pume Trio at 90% Void Break si::e (ft2) Core uncoverv 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.

9PO 9 ow O ~ T 3 o JU_SJ_ i a a -n-346080

Table 2-3. Comparison of System Void Fractions at ESFAS 'Sienal System void fraction ^ Break si::e, (ft ) Punus on Pumos tricoed 2 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 3.45 0.10 2.17 7.97 0.15 10.70 0.20 6.7S D**D aa} Opg-7 1" ? lt A \\ o J0. 5] _1 f( l ' a 1 [- 1 g.,. L r. t>r \\ \\a ,( et '( 1.44i;0Sj_ M LN L L KM f. NUUE. UC bCN L L' L LUN NODE NUMBER DESCRIPTION 1,33 Reactor Vessel, Lever Plenu: 2,34 Reactor Vessel, Core 3,35 Reactor Vessel, Upper Plenum 4,10 Hot Leg Piping 5-7,11-13 Priuary, Steam Generator 8,14 Cold Leg Piping 9,32 Reactor Vessel Devncemer 15 Pressuricer 15,24 Steam Generator Devncener 17,25 Stea= Generator Lc'wcr Plenum 18-20,26-28 Secondary, Steam Generator 21,29 Secas Risers 22,30 Main Steam Piping 23 Turbine 31 Containment MINITRAP2 ?ATH DESCRIPTION PATH NUM3ER DESCRIPTICS 1,2 Core 4'5,46 Core Sypass 3,5,5,11,12,44 Het Leg Piping 6,7,13,14 Pri=ary, stess Generater 8,15 RC Pu=ps 9,16 Cold Leg Piping 10,43 Downe: er, Reactor Vessel 17 Pressurizer Surge Line 18,19 26,27 Steam Generater Downce=er 20,21,28,29 Secondary, Steam Generator 22,30 Aspirator 23,31 Steam Riser 24,32 Stea= Piping 25,33 Turbine Piping 34,35 Break (or Leak) Path 36,37 HFI 38,39,43,44 Ay" 40,41 Main Feed Pu=ps 42 LPI D r3 C3 D') n Table 3.1 cv c; JLl O ~ _ [0] _11' a o 34A;0@ ~

J_ @ i ei i T 2n 1s -t i.

b e

G KD u e lc m G s .s 3 O MNN' d fz w w-n u b Q

1. s p,.

..... - ~.,.......,.. g g W 8' '****8***** 8*da 1 8*da v 2 ' h .s. u.a r sn.. rats a e .,..m...._., 7 2 e, o, i...... m.,

n fath-1.....e. s (ses...me,t Nod e Ne).

Identifteatden Path K'o. Idettifiearfen 1 b oce:er 1,2 C0re 2 Iover Picat= 3,4,1 8,19 Ect 1.cs Pipi:g 3 Core, Core 3ypass Upper

5..' O Ec 1.cc, Upper Plenus, Upper Head 6.21 TO Tubes 50,1c cr Ucad 4,14 Hot k g Piping 7.22 3,15 '

steam Generater Upper 8 Cere mass Eead, >G Tubes (Upper Bal') 9,13,24 Cold Le; Pipins 6.16 5: Tubes (Lover E41f) 10,14,25 h=p s 8,18 50 Lever Head 11,12,13,15.25,27 O ld Leg ?ipics 9.11.19 Cold Lc; Fipi:; (?t=p su::io=) 17,31 ?:vne::er 10,12,20 Cold Leg Piping (?u=p Disetarre) 23 12; 13 Upper :evnec=c: 28,29 Upper tos c::er -(Above the 0 of Sc::.le 3elt) 30 ?rcssuri:c ~ 21 Pressurizer 32 vcat valve 22 contain=ect 33, 3/. w e & rseturn ?a:h 35,36 EPI 27 Comtain=en'. Sprays I o D D ad 1 I Lbi s l r} 8 d),(,(Ib,t> .n

Figure 2-2. CRAFT 2 N0 DING DIAGRAM FOR SMALL BREAKS (6. NODE MODEL) CFT Q 4 2 6 5 3 1 g 1 ~ L LL / /// s m ~~ M@ C LEAK PATHS 8 2, 9 M@ f' e au c=$ Node No. _ jdentification Path No. Identifica:icn 1 ?D Piping, DC, LP 1 Core 2 Pri=a-y SG 2 L?I 3 Core, UP, Mc: Legs 3,10,11 FJI 4 Pressurizer 4 Fo: Legs 5 Containnen: 5 Pu=ps 6 Secondary SG 6 Ven: 'lal ee 7 Pressuri:er 8,9 Leak & 7eturn Ps:h 990 D oo1 py y T g ~ o Ju Jdl ie 34tiOS4

e CORE PRESSURE VS TIME,177 LL 2772 MlVt. PUMPS ON 2 0.023 FT BREAK 23 N00E MODEL c 2 20 0.025 FT BREAK ~~~~ 6 NODE MODEL 18 C2 0 .5 16 E. J E 3 14 2 12 10 qO oo 8 D p 0 500 1000 1500 2000 2500 3 O c3 ~ 7 M ~ @j ~ T.b 1} _ (0 _ k Ju _a 3 Tiine, sec Figure 2-3 2 3 % 065

PERCENT SYSTEM VOIDS VS TIME, PUMPS ON 100 80 a 5 5 60 S o f 1 ( 3 9033 gq '0E I - 40 ~ = reto e o / 4,5 I$. 23' (5 2 f *,# *'/ * 'g. # 20 i / 0 0 400 800 1200 1600 2000 2400 Time, see Figure 2-4 D*O]D L Q o UU ~Q\\ 1f [' O o _fd 8] I_ _a 3460S6

BREAK SPECTRUM-RC PRESSURE WITH THE RC PUMPS OPERATIVE AND 2 HPI PUMPS

2500, 1

2000 n i 1500 J b N d

• ' ' %. - 0 025 FT2 1000 N

\\. N s's \\, \\ -s*%s 500

• g N

~ ' ' -0.05 Fi2 x% ~ 0.075 FT ~.s, ' N. 0.2 FT2 0.10 Fi2 0 0 500 1000 1500 2000 2500 3000 35C Time, se: Figure 2-5 m m~ D D aaS o 'i9 T ~ o 3 o f . Al _1 la 3460S? ~

BREAK SPECTRUM. AVERAGE SYSTEM V010 FRACTION WITH THE RC PUMPS OPERATIVE ANO 2 HPI PUMPS 100 .... w o. 7 o,,, ; m, o m c m ~ s ~ l /* g /',, ~~~~ ~

1. /

,/ G Ci ~%lot r/ "* f*. 80 -~: o / e ,L &j w ? 2i sf W't/ Qsl = o% 60 m- / / \\/ o D/ o + 5 4 / V / s C. i ) /+ / m / 40

• / */

/ % g@ Iy /.. '" /. c. cl / 8 l bc* b d I / ot / c s I v Lv ^ . f v b 20 tC M'S [ ./.- (d'#- v$ (k' g,s 0 0 400 800 1200 1600 2000 2400 2S00 D g g'D Tice, see oo Figure 2 6 ) f aJu_a!_i a . S k5' N

4 2 BREAK WITH CONTINUOUS RC PUMP 0.1 FT OPERATION AND 2 HPI PUMPS 2500 100 g ~ ~ _ __ _ _ _ j LPI

r. /

s sl\\ 1/ k ~ D ]\\ 2000 80 - g s /' M i '"'",' 7 a \\ m3 / p 3 5 6M \\ N 1500 g p Q g, f( L'F-60 I g 3 I \\ g ( / s a a 40 -\\.]' ' 1000 0 m I s, s 3 I s / Ns / N a-20 %'3 500 LPI [ s s% J b.T ' ~ ~ ~~ 0 0 0 500 1000 1500 2000 2500 3100 Time, see Figure 2-7 o r e n,-, f I'.Jjij5.1 4, Sy(gjSg W

RC PRESSURE 'FOR 0.05 FT2 EREAK AVAILABLE 1 HPl VS 2 HPI'S 30 - 2 HPI's, PUMP DN, HOMOGENEDUS 25 o - 1 HPI, PUMP DN, HOMOGENEOUS 20 o ~ .2 E 15 i i S i = i a N ~~~~~ 10 CN e %*% N 0 o %.Ns 0 % 5 50 gg 0 O 500 1000 1500 2000 2500 3000 Time, see Figure 2-B 4 hJ rr 'o 'T p u s]v iII w S 34coso 2 AVERAGE SYSTEM V010 FRACTION FOR 0.05 FT AVAILABLE 1 HPl VS 2 HPI'S 100 o-- -o c-f a f o 80 / / l/ / f

---

2 HP l

• S, PUM? ON,H030 f

E 60 / -o - 1 HP I, PUMP ON, HO"OGE / O f w a0 ,/ = 4 / 2 / E / / 20 0 0 400 800 1200 1600 2000 2400 Time, se: Figure 2 9 DP*D Lo v wy] _ 3 \\1h 1 J0_ j -n-snow.

2 RC PRESSURE FOR 0.075 FT, PUMPS OFF s 905 SYSTEM V010 30 _ _ _ _ 2 HPI'S, PUMP DN, 25 HOMOGENEOUS -. - 2 HPI'S, PUMP 0FF S 90r, V010, 2 PHASE n 20 v .S E 15 d 5 \\ B s _ _ _. _ N 10 sN3 9.sN.N 5 s ~ s % ~ ~ * % ' N. %.% 0 0 500 1000 1500 2000 2500 3000 Time, sec 0** D Figure 2-10

v. v.

id;b .n J _ i o 2 , 3460.9;3

AVERAGE SYSTEM VO!D FRACTION FOR 2 0.075 FT, PUMPS OFF 3 905 SYSTEM V010 ~ 100 / ,r"~-.~. N g,% / 80 f n / ~ / / 2 l 60 / s' 2 / / 2 HPI'S, PUMP DN,H0fi10 GENE 0VS / ~ 40 ,/ 2 HPI'S, PUMP OFF 3 905 V010, = g ,/ 2 PHASE / 20 0 0 400 800 1200 1500 2000 2400 3200 Time, see Figure 2-11 D D L oo. O M ~ 9] ~ L L Ti a JO _ 5 . 3ditiO.<J3

AVAILABLE LIQUl0 VOLUME VS TIME FOR 0.075 FT2 EREAK WITH 1.2 ANS DECAY HEAT CURVE 3000 { 2500 E E 5 2000 3 9' LEVEL OF ACTIVE CORE 1500 l 2 I E l C 0 1000 i, s RC PUMPS '0FF l 500 0 400 800 1200 1600 2000 Time, se: Figure 2-12 D [D a s ip O' ~} ~ m ao a L: m 346[)!)4

2 RC PRESSURE VS TIME FOR 0.05 FT BREAK WITH 1.0 AND 1.2 ANS SEFORE AND AffER PUMP TRIP 2 0.05 FT, 2 HPI'S 3000 1.2 ANS, PUMP DN 2 0.05 FT, 2 HPI'S 1.0 ANS, PUMP DN 2 2500 _, _ 0.05 FT, 2 HPI'S 1.2 ANS, PUMP OFF 2 _ 0.05 F7, 2 HPI'S 0 W. PUMP OFF f._2000 J .a 0 1500 F i 1 ~'

1. ~ 7.,7 s

1000 ,s N.,' * % k [^D' 4i[0;,, 500 s ,q 0 0 500 1000 1500 2000 2500 3000 Time, sec Figure 2-13 ? 91 9 9 i b O) o g-iG:gn 0 iriu t 1 v - 34isOK;

2 PERCENT SYSTEM VOIO FRACTION FOR 0.05 FT BREAK WITH 1.0 AND 1.2 ANS BEFORE AND AFTER PUMP TRIP 100 ' j.>.

==."M N * -8 W,= g /;.Y-s" % 80 ~ m / 60 /-[ 2 0.05 Fi ,.2 HP1'S, PUMP DN, 5 4jf 1.2 ANS ~ = / 2 0.05 FT, 2 HPl*S, PUMP ON, f ~~~ 5 d0 / 1.0 ANS u / 2 2 _ o _ 0.05 FT, 2 HPI'S PUMP OFF, j 1.2-ANS ,/ 2 0 0.05 FT, 2 HPI'S, PUMP 0FF, 1.0 A"S 0 0 400 800 1200 1600 2000 2400 2800 Time, see Figure 2-14 o s TE) ~ 'i9 ~ T f 3 J Ju _ L J a

~ 2 AVERAGE SYSTEM V010 FRACTION VS TIME FOR A 0.075 FT BREAK, BREAK LOCATION COMPARISON PUMPS OFF @ 90G V010

100, UNC0VERY TIME = 625 B0 g

% f/ h /e UNC0VERY TIME = j f 60 f, / g 450 SEC a !/ 40 - g@p,/ g ~@ m g 5 /$g 20 - f</ E 0 0 400 500 1200 1500 2000 2400 Time, see Figure 2 15 D 9O D 6o-og g 7 m a JU J3 L A\\ a - 3 ^., - 34AiO!r;'

COMPARISON OF DEllVERED HIGH PRESSURE INJECTION FLUl0 TO RV FOR PUMP DISCHARGE BREAK \\ 1200 N $, 1100 N <# \\0A 1000 \\#\\/ _ 900 N -1 N = S00 g 700 \\ = '? 505 600 \\ to, = 500 \\ ' Sp, h40G Loop, Hpt \\ ~s 300 N g g LOW togy i SP/ 200 's ~ 100 I 0 0 200 600 1000 1400 1800 2200 2600 3000 Pressure, (psia) Figure 2-16 4 D 9 0 o r9~ wgy,L bd _a w um s 9. m TW l TSV ~ .2 % W (......., e 30 33 g g- ........ - - - - -......... - + s........a I b i MAD,V g ] 7 l }~~' vnov4+ t l (j') + 6 CW2 29 If j jg r--- L@__ 0 j3j i lI e, r,, S y 9, O r i Ol 10 c l I ) @[~ f w,- ".,0 6 q ~ /. If ) ") 6-A7 IL U @r u ~ g-- 3 e /s S 6 s ac u. e y "L J W G --yc I5 ) ) ^ ^ y r; Gl" i ar ' O c-KD t G j17 ,b H 1 ( g 9 15 h e b fe 0 i 1 e e j oo D D Oo O ~] ~ g Jd a (0j J Figure 3.1 MINIrdP2 Nedi:.g and Flow Path Sche =e 34603S

..... c. ! r p T,t i. t ' L. s i J l t. !..... r......,;2 ige.... e.-. e: ere ia. i: t.. .... l 7, c, n..) tito L.;ii >. I i n.. e r ei t.;, v.. u.. 2

e...,.,.. L ! ! c.

r.~-,, ( r <,'i. n., i..<a-e.. s t i,,i t t FP....., D.. Ll.. n.G rT .i i. ( 10 4,- tos r ( RC P!).7 > i d ! P_) Fi:EU ..._._.). t S.G. it'BE RECIC!' a c 3 a fj c 0 FULL l 0 O a> a es n. c PRZR. FULL a w a O E o a" a s a 0 o ~ c-. a e c-r, e e c.>. o 4 b u e O' e 'a r c: c o o - C. o R O C O ,, ri O v e w o o .= a a v -= n v an U E 9 KEY o c 10 6 Q PRESSURlZER O@d o: STEi!* GEi.ERl,T"R ' A ' 6: STEl,M GEN'RATOR '3' 9 i i i 0 2 4 5 S 10 12 14 Trans i ent T ime (" mu tes )' Ficure 3.2 euun u(G NAL 346100

l I ?. i t ..< co t w

c

\\ cr: m a o Ig w < = a. \\ m_ cc ce n. _ C J \\ L:.' teJ LA a m.: :: w n. g D \\ - us w b.- b. [ 6 ( N O O ] M L2 ??.~. : c. &c L, y e T ~ m e < m C'. \\ w w w 6 m g

3 g

\\ c6 m 4 \\ \\ g r= ~ ~ D. Lo M I s \\ m m s w M \\ ( k*s s S. 2= cQ O O < a w s i s k W \\ I g w Z g > - ^ sl s g w s c-a .4 s N s \\ .f4 lg g s \\ a w s s l s G. = m =-m M c s \\ -N D Z \\ 's J W C. C s w \\ g = cc o O \\ \\ \\ \\ 'N g C C # a c: N y c.2 e. - - w a Y,-- y\\ ww r= - y1 w w .a cm :s

= c e m

u. m _ g ' Z [ v C W \\ N e~ \\ss b N o s v. e \\ m t.n c g s g a a w to \\ p-. C.3 gs h 'k % ~ 4 v2 s a = m s \\ s s s ( s Z _Z_ t e e y L3 w N g cc w c: \\ S w cQ 6 't h b N u ,v . w c= a. W D 6 .C in C: II @( C M N W ,"{ } i w N cm C: cm O J f.i i I r l li h l' Ms S #* ~ v e c o O O O G O i c-- - _Z -. g C3 CD cm cm c Lf3 v M N (11) i m i a!alui ;c;;1:u:S n ais ;a:i. ass;;d Figure 3.3 30.ii101

g - -.. - -- s a CO c Z a n f OO c a,, n. c o w c I OO C ( ', y- ' 'l I OO q ~ C- ,e,.0 ~; d I

it OO C

Z a: s m z g cz: o -o I O O G G-

• Z w

w o m c: t w cc C e w - .z l x x w t .Cg g OC C = I ^ w ca => .- m c-: - c: .s o e m o I 3e c a> o u o < a T OC I, "c= W o . m o 0 g=.u c a W a w e" g

s v

l OC m < Cr - o l gg G. n I co

)

w w o LJ a = m .x f Oe c -<m ec - m I O C c c: w w ce c o 22 M o E: l Qm q m hO o E - c l O C = w m" OC G i o o CJ s bC ( l O - v [ OG j ! / ?00R ORIGINE e e ~ l T < I CG / CO G

• ,/

I I m.y. E E o o o o o a o a V M N do) 3)U 213d1'31 1 U'J l 0 03 II(Ul6 ,e* 43 - o A.:. c r '.o d '. t. U 1. J.'.,

a ,..........,. = o-er.. .ws .I 3.- .l...I 45 Cp 9,'.... &*.fl. .iw ) .....J 6 2 (102 FP, S E t:ll. I '. t.: F ! ! f ?. 12 ! FT O nii." E E':0 P'J"T1. E. S T E ' "'. I '.'E C':E??. f "'"'I T l c. 7 r r, ) pq nr-ni. o raio) 7gg PRESSUR!ZER E."P T ! E S L f HOT LEG BEG!t'S TYl0. PHASE YsIN I \\ l (1 g d C' ' FL000 T4.1,K 600 '- s i c I \\ FL0nS LEGINS T ( \\ l \\ i 9 HOT LEG SUSCDCLEO \\ o o o" 5 Q \\ l w N j 09 R\\ u L N =: os c O \\ ci d s 'b 5 i.N ON ss C

b. 400

'h% '~\\ s,/, ~ \\ 8 a s \\ C \\ I O g KEY h \\ Q \\ 0: HOT LEG (PRZR LOOP) 5 \\ be \\ 300 a CORE g a n 6: COLO LEG (PRZR LOCP) 3 SATURATION LINE e ae 200 0 2 4 5 S. 10 Tr:nsient ~ime ("inutu )' Fip,ure 3.E - cm.. ?00R Bl W A

• /. F.D l. v s*

e k CGGL/J.T TE::F~. ii',TU.ES '.'EC. 3"S T'.: .- l E::1 i t '.:: (107' FP BE'il!riikf3 0F LIFE,12 Fi 2 (J..:,LE E;;3 P."I'T"'.'E, 'J::"! T i Cf.TED S TEl"' !!:E D F.E.'. '., P.C FU:.:P Tit lP ) 703 PRESSURIZER EMPTIES I [g yHOTLEG !t BEGli'S TWO-PHASE I hi~ CLO - '\\ CCRE FLOCD TANK l N FLO'.i CEGll'S s o\\ r n D U I HOT LEG SUSC00 LED C a sA o 1 o O s v I a sA 5 000 - o O A s 5 8 u O O Ds o O 0\\ u O O o O s ~ E o g s o O bs/ i u 400 o o o o g O o O o O o O 0 XEY 3 e: HOT LEG (PRIR LOOP) 300 3 G: CORE 6: COLD LEG (PRZR LOCPj o SATURATION LINE r 200 0 2 4 6 3 10 Transient Time ('.h nutes) Figul: 3.5

,5 -

S46104

S T Ll." L U;.'L E V O Lu.c 'n iG US Ti".'.S I E 'T T l; 2 (102 F P ..t 611;ii!.G Gi' L I FE.12.1 F i UL"LE C ' ? f,U ?l :. '.C, U ' f I l !

• l i' 3 S I El "L i n "

12 =- l 9 I. I

n i

11 Ii lI 10 i i SOf hi;%ls:n \\ ' nAl B I t ~ m n i \\ ,- 7 ' w f y l e,. A \\ e e 1 2 l \\ o / I \\ o. ~ y \\ u E / 1 = / \\ ~ ^ l k G I O S 4 \\ \\ Y \\ g w 4 \\ g I I \\ l i f I \\ l \\ \\ g l \\ l i l \\ t [b I 1 2 I s \\ I I i s \\ D t I I lp. a% I \\ / \\ In I N D>s 3 1/ I \\ j#/ sN ti I l g / \\ hk f II Kf,i ' %_M3 __,g, 0 m 0 9 4 6 8 Transient Ti e (inutes) Figure 3.7 0: HOT LEG (PRIR) - RC PUMP TRi? C: HCT LEG '3' LOOP-RC PU?'P TRi? a: HOT LEG (PRIR LOOP) - NO TR!P 0: il0T LEG 'B' LOOP N0 TRIP v4v< A.rsy o s< - so -

e CC' ' Din LE T Fi SI"' V

05 i '. ' "i i l T

.... 1,.9... -r.2 -...,. (,. 0 n -.. F r,...v...'..,..n....

c. e

,r-m,.,. L i. E!.3 h u, i e.E, u... l i l Gai ED Si ti,..L U,2 C;;Et,K) 2500, a r i g. 2000l.- n. i PDM DRillK o 5 1500 - U O0 = 0 6 c. O KEY _u_ E O RC PU"PS Tril? o o a: NO PU" S TP.l? o 1000 C oO u O 6 O o O 500 o o 6 O c3 o oO o6 "a O OaO O o ov o b e i 0 0 2 4 6 3 10 T r a ns i e n t T i.T.e (!.ii a n t e s ) Figure 3.3 d( d o.* g r y,-..UO 8 ..... -.. _... ~ -... - - ~_ -.a a SIEAM GENERATOR AND Pi;ESSURIZER LIQUl0 LEVEL VERSUS TR?.i!SIENT TI',;E 2 DOUBLE E60 RUPiURE -UiCillC.iTED (1G25 FP BEGlHillNG 0F LIFE,12.2 F1 STEAMLINE SREAK) 40 g i O I O I S.G.'S BEGIN TO FILL 7 39 g (NG "'.:P TR IP f u Q CASE O!!LY) "llCa I u C"2 i j. ^ i W 2a A i m E D' c2 e-a j KEY a 0: PRESSURIZER i.EVEL (RC PUMP TRIP) l 6: PFiESSURIZER LEVEL (N0 RC PU;,iP IRIP) l 0: SIEAI.; GENERATOR LEVEL (hD PUi,;P TR!P) O O i a: STLAi4 CENERAT3R LEVEL (RC PU:dP TRIP) b c a i o o a. e 0;{.^_f .g_.;. } Q a a ,,g_c # ^_.2% O l _.e 1 o__ d j 2. 3 K4 0 1 2 3 4 5 G 7 0 9 10 gO Transient Tirac (11:nutes)

e REFERENFES I 3.M. Dunn, et al., "B&N's ECCS Evaluation Model," 3AW-10104, Rev. 3, August 1977. 2 Letter, J.H. Taylor (36W to S. A. '/arga (NRC), July IS,1978. 3 R.A. Hedrick, J.J. Cudlin, and R.C. Foltz, "CFAFT2 - Fortran Program for Digital Simulation of a Multinode Reactor Plant During Loss-of-Coolant," 3AW-10092, Rev. 2, April 1975. 4 J.F. Wilson, R.J. Grenda, and J.F. Patterson, "The 'lelocity of Rising S te. in a Zubbline Two-?hase Mixture," ANS Transactions, 5, (1962). P00R OlBIR e-34.ELO8}}