ML20062A311

Suppl B to Calculation DC-2836, Auxiliary Feedwater Flow Venturi Sizing

ML20062A311
Person / Time
Site:	San Onofre
Issue date:	05/09/1990
From:	SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML13310A269	List: ML13310A268 ML20062A311
References
DC-2836, NUDOCS 9010220005
Download: ML20062A311 (74)

Text

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lg M7 TJCLEAR GENERATION $1TE SYSTEM DESCRIPTION SD.501 620 UNIT 1 REVIS!CN 2 pAGE 2 0F 27 AUX 1LIARY FEEDWATER SYSTEM 1.0 FUNCTIONS / DESIGN BASES 1.1 The Auxiliary Feedwater System has the following main functions:

1.1.1 To provide feedwater to the steam generators during abnormal or emergency conditions which result in a loss of

{

Main Feedwatar.

1.1.2 To provide feedwater to the steam generators during normal start up, normal shutdown and hot stand by conditions.

1.2 The Auxiliary Feedwater System has the following additional functions:

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1.2.1 To provide a means of filling and venting the Main Feedwater System in Modes 4, 5, or 6.

1.2.2 To provide a means of filling and/or feeding the steam generators via the Main Feedwater System in Modes 4, 5 or 6.

1.3 The Auxiliary Feedwater System has the following design bases:

1.3.1 The Auxiliary Feedwater System is designed to provide sufficient steam generator feedwater flow and volume to achieve and maintain the Reactor Coolant System in ' HOT STAND 8Y' (Mode 3) for at least 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, with no offsite power available, following a reactor trip from full power.

The Steam Dump System is used in conjunction with the Auxiliary Feedwater System to meet this design basis.

1.3.2 The Auxiliary Feedwater System is designed such that the Reactor Coolant System can be cooled down to less than 350'F from normal operating conditions with no offsite power available.

1.4 The Auxiliary Feedwater System has the following additional design bases:

1.4.1

%e total minimum delivered flow to the steam genert: ort from either AFW pumps 0 10 and G 105 concurrently) or AFW pump G 10W (opera (operating ting alone) ici i.):

of feedwater events shall be ISS gpm (plus margin) at a steam generator pressure of 1015 psig.

1.4.2 The total minimum delivered flow to the steam generstr.N from AFW pump G 10 for station blackout events shall be W gpm (plus margin [1]) at a steam generator pressure of 10:5 psig.

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This requirement anticipates the future application of generic Station Blackout requirements.-

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REVISION 2 PAGE 2 0F 27 UNIT 1 AUXILIARY FEEDWATER SYSTEM 1.0 FUNCTIONS / DESIGN BASES 1.1 The Auxiliary Feedwater System has the following main functions:

4 To provide feedwater to the steam generators during 1.1.)

abnormal or emergency conditions which result in a loss of Main Feedwater.

1.1.2 To provide feedwater to the steam generators during normal start up, normal shutdown and hot stand.by conditions.

7 i

1.2 The Auxiliary Feedwater System has the following additional functions:

t 1.2.1 To provide a means of filling and venting the Main Feedwater System in Modes 4, 5 or 6.

I 1.2.2 To provide a means of filling and/or feeding the steam generators via the Main Feedwater System in Modes 4, 5 or 6.

t 1.3 The Auxiliary Feedwater System has the fo'. lowing design bases:-

1.3.1 The Auxiliary Feedwater System is designed to provide sufficient steam generator feedwater flow and volume to achieve and maintain the Reactor Coolant System in ' HOT STAN0BY' (Mode 3) for at least 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, with no offsite power available, following a reactor. trip from full power.

The Steam Dump System is used in conjunction with the Auxiliary Feedwater System to meet this design basis.

1.3.2 The Auxiliary Feedwater System is designed such that the Reactor Coolant System can be cooled down to less than f

350*F from normal operating conditions with no offsite power available.

1.4 The Auxiliary Feedwater System has the following additional des 4 n 9

bases:

1.4.1 The total minimum delivered flow to the steam genertter:

from either AFW pumps G-10 and G-105 (operating concurrently) or AFW pump G 10W (operating alone) fci' ix of feedwater events shall be 185 gpm (plus margini at 3 steam generator pressure of 1015 psig.

1.4.2 The total minimum delivered flow to the steam generater.

from AFW pump G 10 for station blackout events shall be 1H gpm (plus margin (1)) at a steam generator pressure of 1015 i

psig.

l

[1]

This requirement anticipates the future application of generic Station Blackout requirements.

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1 NUCLEAR GENERATION $1TE SYSTEM DESCRIPTION S0 501 620 i

UNIT 1 REVISION 2 PAGE 3 0F 37 l

AUXILIARY FEEDWATER SYSTEM 1.0 FUNCTIONS / DESIGN BASES (Continued)

[

1.4.3 The total minimum delivered flow to the unaffected steam generators from AFW pump G-10W for feedwater line break events (upstreamoftheincontainmentcheckvalves)shalf I

be 125 gpa (plus margin) at a steam generator pressure of 1015 psig with operator action to.. equalize flow in each AFW line.

i 1.4.4 The total minimum delivered flow to the unaffected steam 1

generators from AFW pump pumps G 10 and G 105, operating concurrently, for feedwater line break events (upstream of the in containment check valves) shall be 125 gpa (plus margin) at a steam generator pressure of 1015 psig without operator action to equalize flow in each AFW line.

k 1.4.5 The total minimum delivered flow to the unaffected steam generators from AFW pump G 10W or G 105 for feedwater line break events (downstream of the in containment check valves) shall be 250 gpm (plus margin) at depressurized steam generator conditions.

l.4.6 The maximum flow from AFW pump G 105 shall be limited to 420 gpm (pump runout limit) at depressurized-steam generator conditions, considering the most limiting single active failure and using only passive mechanical means.

1.4.7 The maximum automatically delivered flow from AFW pump G 10W or G 105 shall be limited to 150 gpa per steam generator (water hammer limit) at depressurized conditions considering the most limiting single active failure and using only passive mechanical means.

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CASE 4 MainFeedwaterLineBreakDownstreamo(.InContainmentCheck

/N v/f7; Valves at 505 Power l

4a. The plant i,s in'itially operating at 53% of rated power.

l I

4b. Initial' reactor coolant everase temperature is 4'F above the nominal value (551.5'F) corresponding to 50% power level on the nominal average temperature program (575.15'F at full power).

4c. Initial pressurizer water level is 30.0% narrow range span (NRS).

4d. Main feedwater to all steam generators is assun:ed to stop at the time of the feedline break.

4e. Pressurizer power operated relief valves are available, but nn.

l credit is taken for the pressurizer sprays.

4f. AFW is assumed to be manually actuated and the system manually aligned to deliver flow of 225 gpa to two steam generators l

15 minutes after the initiation of the event (feedline break).-

1 49 The steam flow / feed flow mismatch reactor trip is assumed not available..

4h. The feedline break size is assumed to be 0.7854 ft2.. All three i

steam generators depressurize since 50NGS.1 does not have main steamilne isolation valves.

b I

RESULTS CASE 1 Main Feedwater Line Break Upstream of In Containment Check Valves at 100% Power l

The results of the feedline break at ' full power located upstream of'inside containment check valve transient are shown in Figures 1 through 4.

The l

time sequence of events is presented in Table 2.

Reactor trip is provided by the steam flow / feed flow mismatch signal. The results show that an AFW flow of 100 gpa initiated 30 minutes after the break is sufficient to remove core decay heat. Calculations of this case show that the core remained in a coolable geometry during this FLB scenario. The detailed calculations involved showing that the mass relieved through the pressuizar PORVs (between the' time of initial relief through the PORVs and the time the PORVs resent due to the heat removal capability of the AFW exceeding the core decay heat) was not sufficient to uncover the core. As such, the acceptance criterion for a FLB event that the core remains in a l

coolable g30 metry during the transient was shown to be met.

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Main Feedwater Line Break Upstream of In Containment V/d i

CASE 2 at 505 Power M

The results of the feed 1ine break at SM power located upstream of inside l

j containment check valve transient are shown in Figures 5 through 8.

The i

l time sequence of events is presented in Table 3.

Reactor trip is provided by the high pressurizer water level (50% NRS) signal. The results show l

that an AFW flow of 100 gpa initiated 15 minutes after the break is l

sufficient to remove core decay heat. The reactor coolant system (RCS) remains subcooled and the pressurizer does not fill. Thus, the core-i remains covered with water. As such, the acceptance criterion for a FLB event that the core remains in a coolable geometry during the transient l

was shown to be met.

l CASE 3 Main Feedwater Line Break Downstream of In Containment Check l

Valves at 100% Power

~

The results of the feedline break at full power located downstream of l

inside containment check valve transient are shown in Figures g through 12. The time sequence of events is presented in Table 4.

Reactor trip is provided by the steam flow / feed flow mismatch signal. The results show that an AFW flow of 225 gpm initiated 20 minutes after the break is sufficient to remove core decay heat. Calculations of this case show that the core remained in a coolable geometry during this FLB scenario. The detailed calculations involved showing that the mass relieved through the pressuizer PORVs (between the time of initial relief through the PORVs and i

i l

the time the PORVs resent due to the heat removal capability of the AFW exceeding the core decay heat) was not sufficient to uncover the core. As such, the acceptance criterion for a FLB event that the core remains in a coolable geometry during the transient was shown to be met.

CASE 4 Main Feedwater Line Break Downstream of In. Containment Check Valves at 50% Power The results of the feedline break at full power located downstream of inside containment check valve transient are shown in Figures 13 through 16..

The time sequence of events is presented in Table 5.

Reactor trip is provided by the high pressurizer pressure signal. The results show that an AFW flow of 225 gpa initiated 15 minutes after the break is sufficient to remove core decay heat. Calculations of this case show that the core remained in a coolable geometry during this FL8 scenario. The detailed l

calculations involved showing that the mass relieved through the pressuizer PORVs (between the time of-initial relief through the PORVs and the time the PORVs resent due to the heat removal capability of the AFW exceeding the core decay heat) was not sufficient to uncover the core. As such, the acceptance criterion for a FLB event that the core remains in a coolable geometry during the transient was shown to be met.

CONCLUSIONS The reanalysis of the Rupture of a Main Feedwater Pipe supports SONGS 1 l

operation with the reduced AFW flows presented in Table 1.

The reanalysis is applicable for SONGS 1 operation on both the Nominal Tavg Program and Reduced Tavg Program. The radiological consequences following a feedline break were not addressed in this safety evaluation.

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Date c4p/9[.6nt

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Preliminary gi f,

SAN ONDFRE UNIT 1 FEEDLINE BREAK REANALYSIS WITH REDUCED AUXILIARY FEED FLOW AElEMS Due to waterhammer concerns, South'ern California Edison ($CE) is investigating possible modifications to the availiary feedwater (AFW) system. The potential modifications will reduce their (AFW) flow rates.

SCE has requested Westinghouse to reanalyse the feedline break event to support the reduced AFW flows. The feedline break event is the only accident that was reanalyzed.

The previous anaTyses, documented in Reference 1, contains four cases.

Breaks are assumed both upstream and downstream of the in containment 3

check valves initiated when operating at 1005 and 505 of Rated Thomal Power. The analyses documented in this report model only breaks downstream of the in containment check valves. The specific cases that are modeled for this analysis are as follows:

Case 1 - Downstreas FLB initiated at 1035 power with 200 gpa AFW initiated to minutes after the break.

Case 2 Downstream FLB initiated at 535 power with 200 gpa AFW initiated 15 minutes after the break.

Case 3 Downstrees FLB initiated at 1035 power with 175 gpa AFW initiated 20 minutes after the break.

j Case 4 - Downstrees FLB initiated at $35 power with 175 gpa AFW initiated 15 minutes after the break.

Case 5 - Downstrees FLB initiated at 535 power with 30 gpa'AFW initiated 1 minute after the break and increased to 125 gpa 20 minutes after the break.

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CASE 3 Main Feedseter Line Break Downstreas of the In Containment Check Valves at 1005 Power with 175 gpa A N The Case 3 FLB results are shown in Figures e through 12. The time sequence of events is presented in Table 4.

Reactor trip is provided by the steam flow / feed flow mismatch signal. The results show that an A N flow of 175 gpa initiated to minutes after the break is sufficient to remove core decay heat. Calculattens show that the mass relieved through the pressuriser PORVs was not sufficient to uncover the core and thus, the core remained in a coolable gecastry during this FL8 scenario. As such, the acceptance criterten for a FLB event that the core remains in a coelable geometry during the transient was shown to be est.

CASE 4 Main feedwater Line Break Downstrees of the In Containment Check Valves at 505 Power with 175 gpa AFW The results of the Case 4 FL8 are shown in Figures 13 through 16. The time sequence of events is presented in Table 8.

Reacter trip is provided by the high pressurizer water level (805 llRS) signal. The results show that an AFW flow of 175 goe initiated 15 etnotes after the break is sufficient to remove core decay heat. Calculattens show that the mass rolleved through the pressuriser PORVs was not sufficient to uncover the core and thus, the core renstned in a coolable geometry during this FLB j

scenario. As such, the acceptance criterien for a FLB event that the core I

remains in a coelable geometry during the transient was shoun to be est.

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ALL RESULTS ARE OUTPUT EACH PERIOD THIS SYSTEM HAS 52 PIPES WITH 33 JUNCTIONS,

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Pipe Network Test System _

CONTRACT No. 468000 AFW-G-105 IN OPERATION detim PIPE 'NO. NODE NOS.

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JUNCTION NUMBER DEMARD GRADE LINE ELEVATION PRESSURE l

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20

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.00 41.39 31.50 4.28 22

.00 62.40 20.50 18.16 24

.00 57.92 18.80 16.95 25

.00 57.92

. t' 21.70 15.70 26

.00 57.92

'.'22.70 15.26 27

.00 1244.64 16.70 532.11 28

.00 1242.27 15.00 531.82 29

.00 1228.52 15.00 525.86 30

.00 1225.32 21.70 521.57 31

.00 45.58 21.70 10.35 32

.00 41.40 31.50 4.29 33

.00 57.92 18.10 17.26 34

.00 57.92 17.00 17.73 35

.00 57.92 23.15 15.07 36

.00 57.92 23.15 15.07 THE NET SYSTEM DEMAND =

.00

SUMMARY

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• f Nu/re c m o a y. f p4 7e q,p SHBET IEm3 ILNGINEERING DEPARTMENT -

CALCULATION SHEET 0

P-Supplement NtJturlCn SONGS 1 Hydraulie Calentation For AFW t tenaa. Finw Ramiteamaa

CALCNo.,? L RLY 1 J.O.NO.

MADBD oi =

a,"(AP, - api)'

3 A P - A Pr From G10 010S test data:

0,' = Total G10+ G10S flow to S/G's + 3 S/G's 0, = 526 gpm + 3 O, = 175.3 gpm A P, = Po,c,,,,. P.,

6 P, = 1030 psig. 800 psig AP, = 230 psig G.i -

(ns M /AP,-APd' 9

Z E O - 6Pe Loss due to elevation:

From Ref. 14,17,18,19: Z, = 41'. 4 7/16' = 41.37 ft huw oie " 17'

• 0" = 17.0 ft 2,vw oios = 16'.10' = 16.83 ft i

For the largest difference in height, use elevation of Pump 010S.

A = z.,.. z.u..

h,

,41.37 ft.16.83 ft h, = 24.54 ft s.

'",0*b',

l$'c r* b..

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':. ' }_

10. 43. pC1b OPC **

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ett <>n rAu nEv to/s4 (tsvi J

_ _ _ _ _ _ _ _ = _ _ - _ _ _ _ - _ _ _ _ _

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l ENGINEERING DEPAR1NPNI' CALCULATION. SHEET sprunct: SONGS 1 Iivdrnul'c calculatinn For A FW Unas. Finw R = 'rsnents

~

gy- -

_g g j

I 10 NO.

l l

Qi =

(17 5.'s)2 (6P, - lo.G 2') '

i j

i 2 50 - 10.6Z l

i Sample calculation:

Os *

(IM. 33 ( 200 - 10 (oJ1 '

= ' 10 2. 8'7 4PH :

mg

( zso-io.ez)

Q.w 16 7 S'7 X. S = 4 8 8.(o 7074 noW vo - 5/C iS.

3 7

Data to plot System Loss Curve. Refer to Figure 6:

i E Unial Qi.(sain! Q,.Isand 300 201.3 604.0 250 183.1 549.4 200 162.9 488.6 180 154.0 462.1 160 144.7 433.9 r

i 140 134.6 403.9 100 111.9 335.7 60 83.2 249.5 20 36.2 108.6 10.62 0

0 System curn can be validated with the 010 test data for FCV's fuity open with S/O pressure at 800 psig.

Refst to Figure 6 for test data points.

I e

a Y

a SCl! On 3"in tun ieju (cw)

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........