Forwards Suppl B to Calculation DC-2836, 'Auxiliary ...

Calculation Title Page

Project -56 FJG7 Job Order No. Discipline M5C14

Subject A re-,/ov V 6 ./-r v -r S i & ci 6A, 5U P0#4. a

Calculation No. 0 4- 2S 3C QA Class No. Pages 0

Responsible Engineer (h . Date.31 L-L....

Independent Review Engineer--/,; ' Date SIGNATUNW

ORIGINAL ISSUE

NAME DATE SIGNATURE

Group Leader K. A TTE/ L

Discipline Sup. Engineer /A - WAZ7_70A_

Professional Engineer

(if required)

RECORD OF REVISIONS

NO. REASON FOR REVISION DATE RESP. ENGR. IRE GL DSE PE

.10 Olz 00

00n 0 0

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NUCLEAR GENERATION SITE SYSTEM DESCRIPTION 50-501-620 < UNUA T SIT REVISION 2 PAGE 2 OF 37

AUXILIARY 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 Feedwater.

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:

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 STANDBY' (Mode 3) for at least 32 hours, 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:

I..1 UThe total minimum delivered flow to the steam cenerator from either AFW pumps G-10 and G-10S (operating concurrently) or AFW pump G-IOW (operating alone) fc : of feedwater events shall be 185 gpm (plus margin) at a steam generator pressure of 1015 psig.

1.4.2 The total minimum delivered flow to the stear generato from AFW pump G-10 for station blackout events shall be 135

gpm (plus margin (1]) at a steam generator pressure of .015 psig.

[1] This requirement anticipates the future appication a generic Station Blackout requirements.

NUCLEAR GENERATION SITE SYSTEM DESCRIPTION S0-SO1-620 UNIT 1 REVISION 2 PAGE 3 OF 37

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 (upstream of the in-containment check valves) shail be 125 gpm (plus margin) at a steam generator pressure of 1015 psig with operator action to equalize flow in each AFW line.

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

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

1.4.6 The maximum flow from AFW pump G-1OS 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-10S shall be limited to 150 gpm per steam generator (water hammer limit) at depressurized conditions considering the most limiting single active failure and using only passive mechanical means.

57 F -5 f5 cF 61 CASE 4 Main Feedwater Line Break Downstream of.In-Containment Check

Valves at 50% Power V 4a. The plant is initially operating at 53% of rated power.

4b. Initial reactor coolant average temperature is 4'F above the nominal value (551.50F) 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 assumed to stop at the time of the feedline break.

4e. Pressurizer power-operated relief valves are available, but no 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 gpm to two steam generators 15 minutes after the initiation of the event (feedline break).

4g. 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 steam generators depressurize since SONGS-1 does not have main steamline isolation valves.

RESULTS

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

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 e 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 gpm 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 pressuizer PORVs (between the time of initial relief through the PORVs and the time the PORVs reseat 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 gqometry during the.transient was shown to be met.

0 AAOE B If- 4/ , ifC..4 / OH 7

ATTAcflM45A 7-F~~

CASE 2 Main Feedwater Line Break Upstream of In-Containment Check Valves at 50% Power +f3

The results of the feedline break at 50% power located upstream of inside containment check valve transient are shown in Figures 5 through 8. The 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 that an AFW flow of 100 gpm initiated 15 minutes after the break is sufficient to remove core decay heat. The reactor coolant system (RCS) remains subcooled and the pressurizer does not fill. Thus, the core remains covered with water. As such, the acceptance criterion for a FLS event that the core remains in a coolable geometry during the transient was shown to be met.

CASE 3 Main Feedwater Line Break Downstream of In-Containment Check Valves at 100% Power

The results of the feedline break at full power located downstream of inside containment check valve transient are shown in Figures 9 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 the time the PORVs reseat 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 gpm 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 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 the time the PORVs reseat 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 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.

AAADE 3/ A/

D47-

Made By Date ' Calc No.D04S97 Rev.

kBDt (f ent Transmittal PA ; Sh 95-S'2 of By By Date ~ - r ~J~B* Preliminary j 14

SAN ONOFRE UNIT 1 FEEDLINE BREAK REANALYSIS WITH REDUCED AUXILIARY FEED FLOW

BACKGROUND

Due to waterhammer concerns, Southern California Edison (SCE) is investigating possible modifications to the auxiliary feedwater (AFW) system. The potential modifications will reduce their (AFW) flow rates. SCE has requested Westinghouse to reanalyze the feedline break event to support the reduced AFW flows. The feedline break event is the only accident that was reanalyzed.

The previous analyses, documented in Reference 1, contains four cases. Breaks are assumed both upstream and downstream of the in-containment check valves initiated when operating at 100% and 50% of Rated Thermal 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 I - Downstream FLB initiated at 103% power with 200 gpm AFW initiated 20 minutes after the break.

:Case 2 - Downstream FLB initiated at 53% power with 200 gpm AFW initiated 15 minutes after the break.

Case 3 - Downstream FLB initiated at 103% power with 175 gpm AFW initiated 20 minutes after the break.

Case 4 - Downstream FL8 initiated at 53% power with 175 gpu AFW initiated 15 minutes after the break.

Case 5 - Downstream FLB initiated at 53% power with 30 gpa AFW initiated 1 minute after, the break and increased to 125 gpa 20 minutes after the break.

ATTA CH1M EV T E CKEF L)

Made By Date Calc No.Oct.34 Rev. 0

Ckd By Date -- .... - - m tSh - of t rnna int Tr~anarnittal1 *

Westinghouse Proprietary Class 2 Preliminary

CASE 3 Main Feedwater Line Break Downstream of the In-Containment Check Valves at 100% Power with 175 gpm AFM

The Case 3 FLB results are shown in Figures 9 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 175 gpm initiated 20 minutes after the break is sufficient to remove core decay heat. Calculations show that the mass relieved through the pressurizer PORVs was not sufficient to uncover the core and thus, the core remained in a coolable geometry during this FLB scenario. 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 the In-Containment Check Valves at 50% Power with 175 gpm AFW

The results of the Case 4 FLB 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 water level (50% NRS) signal. The results show that an AFW flow of 175 gpa initiated 15 minutes after the break is sufficient to remove core decay heat. Calculations show that the mass relieved through the pressurizer PORVs was not sufficient to uncover the core and thus, the core remained in a coolable geometry during this FLB scenario. 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.

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I. IZS.tO 72.77 538.1- 0.744 2.33.714 517.$ 304._ 15. 1a8.S7 7Z.61 537.55 0.5318 147 V7.5 5701 1-o.Y.0

-4D G _ _ -92-3/1 C_ _ s,

- .P f L a 0C c.93 Go ACV R 0'v

A~O T-F 7-.) 3/z.M E CH1Z5-,r7, ' .2 2r~ 3 L)

28 .00 15.00 41 42 43 2834 29 .00 15.00 .42 43 44 P 30 .00 21.70 44 45 46 P 0OF P.IS 31 .00 21.70 45 46 47 BY Mt0cAs(.-vDATE(-6-f 32 .00 31.50 47 48 33 .00 18.10 34 35 49 CHK BY DATE I 34 .00 17.00 35 49 50 35 .00 23.15 50 51 36 .00 23.15 51 52

OUTPUT SELECTION: ALL RESULTS ARE OUTPUT EACH PERIOD

THIS SYSTEM HAS 52 PIPES WITH 33 JUNCTIONS , 11 LOOPS AND 9 FGNS

THE RESULTS ARE OBTAINED AFTER 4 TRIALS WITH AN ACCURACY = .00027

Pipe Network Test System CONTRACT NO. 468000 AFW-G-10S IN OPERATION

PIPE NO. NODE NOS. FLOWRATE HEAD LOSS PUMP HEAD MINOR LOSS VELOCITY HL/1000 LINE 1 IS CLOSED

2 2 11 .00 .00 .00 .00 .00 .00 3 11 0 .00 .00 .00 .00 .00 .00 4 2 3 .00 .00 .00 .00 .00 .00 5 3 4 .00 .00 .00 .00 .00 .00

THE CHECK VALVE IN LINE NUMBER 6 IS CLOSED THE CHECK VALVE IN LINE NUMBER 7 IS CLOSED THE CHECK VALVE IN LINE NUMBER 8 IS CLOSED 9 5 6 131.13 3.70 .00 .00 5.68 39.38

10 6 7 65.56 13.75 .00 .00 2.84 10.72 11 6 7 65.56 13.75 .00 .00 2.84 10.72 12 7 a 131.13 3.07 .00 .00 5.68 39.38 13 8 9 131.13 .32 .00 1178.42 5.68 39.38

LINE 14 IS CLOSED 15 9 10 131.13 4.02 .00 .00 5.68 39.38 16 10 0 131.13 .02 .00 .00 .56 .13

THE PUMP IN LINE 17 IS OPERATING OUT OF RANGE q LJOS)tjv l- 4 US.i ) 17 0 13 405.45 2.18 1296.30 .00 17.57 345.81 18 13 22 12.00 1248.93 .00 .00 4.44 94.95 19 22 0 12.00 4.48 .00 .00 1.89 11.38 20 13 14 393.45 47.62 .00 .00 17.05 326.15 21 14 15 393.45 18.15 .00 .00 9.90 81.38 22 15 16 -- 131.14 .87 .00 .00 5.68 39.39 23 15 5 131.13 .52 .00 .00 3.65 13.02 24 15 27 131.19 .55 .00 .00 3.65 13.03 25 16 17 131.14 2.32 .00 .00 5.68 39.39 26 17 18 65.56 13.77 .00 .00 2.84 10.72 27 17 18 65.58 13.77 .00 .00 2.84 10.73 28 18 19 131.14 . .00 .00 5.68 39.39 29 19 20 131.14 .32 .00 1178.61 5.68 39.39

LINE 30 IS CLOSED---31 20 21 131.14 4.65 .00 .00 5.68 39.39 32 0 131.14 .02 .00 .00 .56 .13

34 24 33 .00 .00 .00 .0 Y or; JoDATE\ 00o LINE 35 IS CLOSED CHK

36 24 25 .00 .00 .00 .0.000 37 25 26 .00 .00 .00 .00 .00 .00 THE CHECK VALVE IN LINE NUMBER 38 IS CLOSED THE CHECK VALVE IN LINE NUMBER 39 IS CLOSED THE CHECK VALVE IN LINE NUMBER 40 IS CLOSED 41 27 28 131.19 2.36 .00 .00 5.69 39.42 42 28 29 65.59 13.76 .00 .00 2.84 .10.73 43 28 29 65.59 13.76 .00 .00 2.84 10.73 44 29 30 131.19 3.19 .00 .00 5.69 39.42 45 30 31 131.19 .32 .00 1179.43 5.69 39.42

LINE 4N6 IS CLOSED 47 31 32 131.19 4.18 .00 .00 5.69 39.42 48 32 0 131.19 .03 .00 .00 .56 .13 49 33 34 .00 .00 .00 .00 .00 .00 50 34 35 .00 .00 .00 .00 .00 .00 51 35 36 .00 .00 .00 .00 .00 .00 52 36 0 .00 .00 .00 .00 .00 .00

JUNCTION NUMBER DEMARD GRADE LINE ELEVATION PRESSURE 2 .00 57.92 17.00 17.73. 3 .00 57.92 16.00 18.17 4 .00 57.92 16.70 17.86 5 .00 1244.66 18.30 531.42 0 6 .00 1240.96 15.75 530.93 7 .00 1227.22 15.75 524.97 8 .00 1224.14 23.20 520.41 9 .00 45.40 23.20 9.62

10 .00 41.39. 31.50 4.28 11 .00 57.92 21.80 15.65 13 .00 1310.95 16.80 560.80 14 .00 1263.33 16.80 540.16 15 .00 1245.18 16.70 532.34 16 .00 1244.32 15.00 532.70 17 .00 1241.99 15.00 531.70 18 .00 1228.23 15.00 525.73 19 .00 1224.96 ..22.40 521.11 20 .00 46.04 22.40 10.24 21 .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 ! 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 OF INFLOWS (+) AND OUTFLOWS (-) FROM FIXED GRADE NODES

1b 01

CL ebv,.

iO.Zrl.

to MAeIOIA-C 3Eg

2~~ 131f.I5

<i>4

qDSBY i03A Ac

17K 17 81 AE z

CF'<.A £Lh A£-)-i 140'

D = = ---- -- - - -+-

-1I rr -

- , I

T-1 r-i

Li~ rJ

-:7

-~ T

---- --- -- -- b L Trf D -F -ELL;w~oJ b~ _~ .+:17 ~ Ph

__

F V ~ , ,

______~ -4

TOTAL LOW __

loo - fi f ~ rza 'FTuq

SHEET1 IEEPS ENGINEERING DEPARTMENT

CALCULATION SHEET 4- . . Supplement

Duniiact SONGS 1 Hydraulic Calculation For AFW Lines - Flow Reqpkrments CALC NO. REV. 3

J.o.NO. MADE D

L\ ~P, -\PE

From G10+G10S test data:

Q2 Total G10+G10S flow to S/G's + 3 S/G's

0 2 = 526 gpm + 3

Q2 = 1753 gpm

-AP 2 = POISCHARGE - S/O

L P2 = 1030 psig - 800 psig

t P,= 230 psig

Loss due to elevation:

From Ref. 14, 17, 18, 19: Za = 41' - 4 7/16" = 41.37 ft

ZPuuP sio = 17' - 0" = 17.0 ft

Puu6P.1os = 16' - 10" = 16.83 ft

For the largest difference in height, use elevation of Pump G10S.

hL ' Z;/ "PUMP GUos

b = 41.37 ft - 16.83 ft

hL = 24.54 ft

C ) 3on v REV 10/84 (CWE)

A 06)I~T f.c ~

ENGINEERING DEPARTMEI

CALCULATION SHEET suw~r SOGS 11-ldrauic Clcuition For AFRV Line., - Flow Rccndrements j me t

J.O.NO. A_

17 53) P , o2.) F7I~s.Z BOA - O .toz*

Sample calculation:

(15.' z7 S20 0l 2. o2. GZ 87 PM

1L ,4 8,, G TO-rAL F:LOW T-O

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

m~idy Q,.LrnQ Gral

300 201.3 604.0 250 .183.1 549.4 200 162.9 488.6 180 154.0 462.1 160 144.7 433.9 140 134.6 403.9 100 111.9 335.7 60 83.2 249.5 20 36.2 108.6 10.62 0 0

System curve can be validated with the GIO test data for FCV's fully open with S/G pressure at 800 psig. Refer to Figure 6 for tcst data points.

4C0 ~37-~RV;O~ C.

Forwards Suppl B to Calculation DC-2836, 'Auxiliary ...

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