ML20148B234

Responds to NRC 791221 Request for Info Re Applicability of Current Analysis for Main Steam Line Break.Forwards Analysis & Evaluation of Automatic Initiation of Auxiliary Feedwater Sys

ML20148B234
Person / Time
Site:	Fort Calhoun
Issue date:	01/10/1980
From:	William Jones OMAHA PUBLIC POWER DISTRICT
To:	Reid R Office of Nuclear Reactor Regulation
References
NUDOCS 8001210261
Download: ML20148B234 (45)

Text

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.j Omaha Public Power District 1623 HARNEY a OMAHA, NEBRA8MA 68102 s TELEPHONE S36 4000 AREA CODE 402 January 10, 1980 Director of Nuclear Reactor Regulation ATTN:

Mr. Robert W. Reid, Chief Operating Reactors Branch No h U. S. Nuclear Regulatory Commission Washington, D. C.

20555

References:

(1) Docket No. 50-285 (2) NRC Letter from Robert W. Reid to W. C. Jones Regarding Automatic Initiation of Auxiliary Feedwater Systems at Fort Calhoun, Dated December 21, 1979 Gentlemen:

The referenced letter requested Fort Calhoun response to a con-cern regarding the applicability of the current analysis for main steam line break and requested Fort Calhoun to resolve this concern by submitting an analysis within twenty (20) days from the receipt of that letter.

The attachments to this letter provide the NRC staff with suf-ficient information to resolve this concern. The attachments to this letter include the following:

1.

Best Estimate MSLB Analysis to Assess NSSS and Contain-ment Respense With Automatic Auxiliary Feedwater Actuation 2.

Evaluation of the Impact of Automatic Initiation of Auxi-liary Feedwater on MSLB Analysis Based on the best estimate analysis provided, it is our belief that the control grade system proposed presents a safe means of operation.

It is also oar belief that current means of manual actuation of auxiliary feedwater is also safe. We do not share your opinion that the. questions in the referenced letter are applic-able to the manual mode of operation, since operators are directed to isolate the affected steam generator on a steam line break.

In addition to the best estimate analysis, the effect of auto-matica11y initiating auxiliary feedwater during MSLB vas evaluated using licensing assumptions. The evaluation was based upon calcula-tions computing, return to power. The conclusions of our evaluations

,4 0 @

s 8001210tol*

a 95024344 s

. are that, with a three minute delay in the actuation of auxiliary feedvater, the conclusions of the MSLB events presented in the FSAR and subsequent reload submittals conservatively bound those with control grade automatic auxiliary feedvater system for return to power.

Sincerely,

/

y' W. C. Jones i

Division Manager Production Operations WCJ/KJM/BJH:jmm Attach.

cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, H. W.

Washington, D. C.

20036 i

911024345

Enclosure (2) a to 4'- 'g[*

CE-18074-706 4

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4 BEST ESTItiATE liSLB AtlALYSIS TO ASSESS NSSS

^

AND CONTAINiiENT RESP 0flSE WITH AUT0f1ATIC AUXILIARY FEEDWATER ACTUATI0ft e

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8 96024.546

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i.0 tillR000CT10f t A set of calculations has been per,smed on a generic basis with plant i

i characteristics representative of CE operating plants to nodel containment building pressure and temperature response and overall flSSS behavior.

including core reactivity, following a tiain Steam Line Break (11SLB) inside l

containment. The intent of these calculations is to determine if the containment building response (pressure) and the core reactivity response (return to power) are acceptable following a (1SLB when auxiliary feed-water is added without regard to the identification of the affected stean generator. The auxiliary feedwater flow is assumed to be activated at the initiation of the transient to naximize its e'ffects. liain feedwater flow including post trip rampdown is simulated. fio isolation of main or auxiliary feedwater is considered unless a high water icvel condition is reached.

2.0 ASSUtiPTI0flS At1D CASES Assunptions for the analyses are given-in Table 1.

The four cases analyzed are listed in Table 2.

i 3.0 ' DISCUSS 10!! 0F RESULTS

!!aximum containment pressure and least negative core reactivity for the four cases are listed in Table 3.

Both the containment pressure and the reactivity (return to power) values are. within acceptable limits.

liain feedwater flow, auxiliary feedwater flow, core reactivity change, core power, containment pressure, primary loop temperatures, and steam' generator secondary temperatures for the four cases are detailed in Figures A-1 through A-7, B-1 through B-7, C-1 through C-7, and D-1 through D-7, respectively.

The results of the analyses using best estimate nodels for steam generator moisture carryover and containnent passive h' eat sink heat transfer demonstrate that the additional auxiliary feedwater has a negligible impact on containment peak pressure. The containment peak pressure is detemined primarily by the initial inventory in the ruptured unit. This 1

98024347

inventory is released within the first few minutes, depending upon the break size, so that the contribution of auxiliary feedwater flow to the ruptured unit over this time frame is small. Over the longer time franc, the secondary inventory is boiled off at essentially the decay heat rate which the containment active heat removal systens can accor.nodate while reducing containment pressure. The excess feedwater which is not boiled off remains in the steam generator, causing the secondary level to ri.se.

The containment peak pressure is es.sentia11y an initial inventory limited phenomenon.

The results of the analyses also show that th,e additional auxiliary feed-water has a negligible impact on core reactivity. Cases A and C assuno no stuck rods and a best estimate moderator cooldown curve.

For conparison, Cases B and D assune that the most reactive rod is stuck and that the moderator cooldown curve is a licensing curve. All cases took credit for boron injection via three charging pumps; however, safety injection boron credit was not taken. Those cases do not have a return to power for the following. reason. The initial primary loop temperature decreases are linited 2

by' the two-phase blowdown process associated with large break P2 ft),

since much of the break flow is saturated liquid which has not absorbed significant amounts of energy from the primary loop.

For smaller break 2

areas (c2 ft ), the blowdown is pure steam which does require large amounts of energy per unit mass to boil via primary to secondary heat transfer; however, the rate of primary-to-secondary heat transfer is controlled by the blowdown flowrate which in turn is limited by the small break area. The net result is that over approximately the first 100 seconds of the event, the amount of core and loop cooldown is about the same regardless of break size. This time frame is most important since.the presence of delayed neutrons minimizes the amount of cooldown needed to produce a core criticality problem.

9 Without a return to power (via primary loop cooldown and delayed neutrons),

the remainder of the transient is a gradual increase in reactivity due to loop cooldown which is coupled to the containment pressure, plus a decrease in reactivity due to boron injection In time (approximately 300 seconds),

the reactivity decrease due to boration overtakes the reactivity increases due to loop cooldown; thereafter, the total reactivity steadily decreases.

The ~ ruptured' steam generator is at the containment backpressure and with 96024548

RCPs operating the sensible heat from the non-ruptured unit is quickly removed resulting in RCS and SG secondarv temperatures essentially in equilibrium with the containment conditions in about 10 minutes.

liith licensing assunptions, the peak in the reactivity transient _is calculated to be within the first two minutes of the event. A three minute time delay, if added to the automatic actuation circuit, would justify a statement that automatic auxiliary feedwater actuation will not inpact existing SAR core cooldown MSLB analyses.

4.0 COMPARIS0l1 WITH LICEllSIllG CALCULATIOlls The following items are important in comparing the results contained herein with those obtained with traditional licensing models and assumptions:

1.

The moisture carryover model used is a best estimate model which gives a two-phase blowdown for large break areas. The two-phase blowdown results in a lower containment pressure and less initial primary loop cooldown than a pure steam blowdown. Chapter 15 analyses assume a pure steam blowdown regardless of break size.

2.

Chapter 15 analyses assume that the most reactive rod is stack.

lloreover, the remaining rod worth is assigned a conservative value in conjunction with a conservative moderator cooldown curve.

3.

A best estimate containment heat transfer medel provides containment pressurization results significantly lower than those provided in Chapter'6 analyses.

4 0

99624349

llujLt i e

ASSUttPTIO!!S

[15S5 Initial Conditions

• 2700 flut Power Core Inlet Temperature 548'F Primary Pressure 2250 PSIA Secondary Pressure 875 PSIA Secondary Temperature 529'F Containment Data 6

3 Free Volume 2.5 x 10 ft Design Pressure 44 psig Heat Sinks SAR values Heat Transfer Model Best estinate model Number of Fan Cod ers 4 (no single failure) 6 Fan Cooler Capacity, each 68 x 10 B/hr at 280'F containment tenperature 100 F CCW Temperature Fan Cdoler Actuation Setpoint Fans are operational 0t=0 Number of Sprays 2(nosinglefailure)

Spray rate, each 2700 GPl1 Spray Actuation Setpoint 10 PSIG + 60 seconds

~

Other Data Steam Generator Isolation Signal (MSIS)setpoint 500 psia Decay Heat Curve NIS-5 Main Feedwater Flow o

1 Ruptured Unit:

Ramped to 10!; over 60 secondt following Reactor Trio: (10'.

represents twice the bypass nominal value of 5".. this accounts for pur.o run-out wit reduced backoressure),

temperature is reduced to 100 to account for turbino off.

line.

Flow terrinated if th;

~

clovation of ucser level tao is reached.

See figures A-1' B-1 c-1

^"d -'-

9024350

TABLE 1 ---- continued liain Feedwater Flow -- continued Unaffected Unit:

Same as ruptured unit except that flow is ramped to 5%.

See Figures A-1, B-1, C-1 and 0-1.".

Auxiliary Feedwater Flow Ruptured Unit:

Mitiated at t = 0.

Flow rate is r function of unit pressure. All a;ntrol valves assumed to be fully opened.

Unaffected Unit:

flo flow; all flow is totally diverted to the ruptured unit.

Reactor Coolant Pumps Operating during the transient.

CEA Insertion !! orth All rods in (ARI)

-B.9%

(nostuckrod) liost reactive rod stuck

-7.12%

(bestestimate) tioderator llorth SAR Value See Figure 1 Best Estimate klue See Figure 2 Doppler llorth See Figure 3 lioisture Carryover On Steam Genera. tor Secondary Side Best Estir. ate liodel Boron Injection Parameters

~

~

Safety Injection Credit !!ot Taken Charging Pumps llumber of Pumps 3

Flow Rate 44 GP!! per pump Actuation Time SIAS 8f,by weight Doric Acid Concentration Doron llorth 00 PP!1/%

Boric Acid Conversion Factor 1749 PPf1 boron /i; by weiDht boric I

liixing tiodel Used Slug Flow Itodel Loop Transit Time 10.5 seconds 96024351 ess e

o TABLE 2 CASES Case CEA Scram Worth (,)

tioderator Curve Dreak Area (Ft )

A

-8.9 Figure 2 6.63(I)

II)

B

.-7.12,

Figure 1 6.63 C

-0.9 Figure 2 1.99(2)

D

-7.12 Figure 1 1.99(2)

(1) Double-ended severance of main steam line (two-phase blowdown).

(2) Largest break area corresponding to pure steam blowdown.

+

1

)

96024352

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.....~~...----~;-----L-._,,_-----.-~~-,,.---------,_,,_...,-

a' 5

' TABLE.3

'RESULTS

....t... '

i

., L,iast-flega'ti,ve '

Case Containment Peak Pressure (PSIG)

Core Reactivit v",

A 29.7/83.0(soc.)

-4.31 B

29.7/83.0(sec.)

-2.34 C

35.0/231.9 (sec.)

-3.54 D

35.0/231.9 (sec.)

-1.55 t

e I

4 96024553

FIGURE 1 REACTIVITY VS'liODERATOR TEl1PERATURE 5

(SAP.VALUE) 3 q

4 is 4 z

O

- p x

W

.m 3.

z H

h m

~

>--o2 w

x 1

0

-1 i!

i i

i i

300 350 400 450 500 550 600 L.,O D m.o.:. imt..o, T:...c :.a.a.... =.=, 0 =

c s

s v.s e

97024554

7 s

s s

a s

FIGURE 2 PEACTIVITY VS 110DERATOR TE!!PEPATURE g

(BEST ESTl!! ATE VALUE) 5

.~

s s

Q.

4

.ts 4

~

z o

W x

w 3.

m se m

s H

m no2 w

x t

1

\\.

1 300 350 400 450 500 550 603 M0 DER.iTOR TE.'l. PERI.T URE, OF

'c)g)()24 355

FIGURE 3 1,g.l g

DOPPLER REACTIVITY VS FUEL TE!'.PERATURE I

3,g s 1.4

.f s

.. i

.1.2 -

1.0 i

.g,g.

0.6 i.

\\

y, n,4 i

.c, F-.

x w

(.o

=

0.0

~

500 1000 1500 2000 2500

.r_,

4......

u w

cc 1,

i.... -.....

-0.G -

.w

.... 0.

t 1.0 i

\\

4 n

.t I

i 96024556 FUEL TEfIPERATURE *F

~

\\

i

..s

i J

1800 1600 u

FIGURE A-1 w

v)

N 5

1400 11AIff FEEDilATER FL0kl VS TIllE

-J zo;'

1200 ccw s-

.c 1ow W

100C u.

z cr

\\

800 4

600 t

400 200 AFFECTED U;llT UtlAFF,ECTED Ul!IT O

I I

I I

. 6m. o 0

200 400 600 80g

,rnn Tlt5b',!SICI

.e 96024357

360 f

FIGURE A-2 320 Ef1ERGEllCY FEEDWATER FLOW VS TIf1E 110 FL0ll TO U:lAFFECTED U: LIT u

w en s

r w

d 240

=

0 J

LL cew 200 s

a cr 2

Ow W

tu 160 3uzw i

a cc w

v

[L 120 80 l AO L 98024358 O

I I

i t

i 0

200 400 600 800 1000 160C TIME ISEC)

FIGURE.A-3 REACTIVITY CilA'lGES 3

VS.

8

'TiltE e

~

!!ODERATOR 4

~

2 sz LLi u

M lF DOPPLER W

r n

0 s

~>

>-o cc

-2 w

BOR0tl cc i

-4 TOTAL i.

-6 il is\\

-6..

95024559 CEA

-10 I

._.I

_____.L _

1 J

0 200 400 600 S00 1000 120C TINE fSEC1

t 0.D i

FIGURE A-4 0.80 CORE POWER VS TIltE 0.70 0.60 0.50 uJ 2o a.

w M

0.4 0 o

U 0 l._

0.3 l

0.2 0 0.1 0

~

98024360 0.0 0 __

t I

I l-l 0

200 400 600 800 1000 120; l

T I!1E CSE C )

r-FIGURE A-5 55 -

CotiTAIrltiErlT PRESSURE VS TIl1E

~

50 45 c

f 40

-we aww w

35 x

c_

Z w

rz 30 c

s

--g u

25 l._

t 20 i

o 15

%025001 t

10 I

I 0

200 400 600 800 1000 120:

TIME (SEC)

F r

e

's

\\

FIGURE A-6

. PRIfMRY LOOP TEllPERATURES 540 vs I;'

.TIllE p

480 COLD LEG OF Ull.A..F..F.EC..TED Utl.IT. '

u.

-110T LEG

~

.s 360 a

wo u)

RJ 5

30Q COLD LEG OF AFFECTED.UtlIT.'

H g

trw Q

r

• Ll H

2.40 c.

oo J

1 180 s.

120 60

~

90025002 l

O f

I f

I l

0 200 400 600 800

!Cou-120; re.e,ceci

.4 FIGURE A-7 540 STEAf1 GEllEMTOR TE!!PEMTURES '

vs TIllE 480 420 u.

O w

U UllAFFECTED UilIT 360 weo F--

C Zw c

^ ~

si 300 w

i e

AFFECTED UtlIT o

F--

C M

240 w

7-We r

C*

~180 to 1

120 d

60 90025003 0

I L

1 I

I O

200 400 600 800 1000 120:

TIME ISEC1 i

....:..~.._..:.----.---.---.-.-....-.:--......-..

.. -. _. =. _. = _.. _. _.

I

.)

1800 1600 4

I w

FIGURE a-1 v>

N E

1400 11Alli FEED!!ATER a

FL0t! VS TillE 2

O J

LL.

1200 te

\\

w

+_

c 3

j a

w W

100C g

z

~

Cz 800 f

600 t

400

~~

M025004 200 '-

AFFECTED U;11T i

UtlAFF,ECTED U!:IT l

0 i

i j

J l

0

~ 200 400 600 800 1000

200 I

TIME ! SE.C ;

360 _'

I FIGUP.E B-2 320 E!!ERGEllCY FEEDUATEP, FLOW VS TIttE fl0 FLOW TO UtlAFFECTED 28n UtlIT u

w vu s,

E co d

240 1o a

' LL ew 200 T

1a www 160 3

uz w

s e

g w

5 120._

t 80._

40.

J 90025005 0

t I

0 200 400 600 800 1000 120';

TIME ISEC1 9e e,

,-,g-

l.

s FIGURE B-3 REACTIVITY CHAllGES i

y5 0

TIllE 6

110DERATOR 4

e 0

[

2 y

z i

w u

DOPPLER D

I n

ee Hu C

-2 1

g e

TOTAL TOTAL

-4 f

i BOR0tl 2

l

-6.

l CEA

-8 90025006

-10

.L 1

0 2r;0 400' 600 600 1000 120C Tir1E (SEC)

FIGUNE3-4 a

0.90

. cone P0llEn VS

, TI!!E 0.00 0.70 P

0.60 0

0.5 u

e w

2 o

c.

w e

0.40 i

o U

0.30 a 0

0.2 0.10 L c0025007 000__

~

t i

i 0

200 400 600 800 1000 l200 TIl1E.(SEC1

FIGURE B-5 c.

55 C0tlTAlll!!EllT PRESSURE

, VS' TIllE 50 i.

45 N

c m

40 Q_

i we om m

w 35 e

D.

H z

wz z

30

\\

~c W

z a

u 25 t

20 15

%025008

~

10 1

1 I

I 1

0 200 400 600 800 1000 120'j TINE (SEC1

t j

FIGURE,B-6 PBIllARY LOOP TEftPERATURES 540 b i.vs i

I

.TIltE

- f s

l

~

~

480 COLD LEG OF U.flAFFECTED.UtJIT 420 HOT LEG <

~

u.

360 a

w a

to W

eo 300 COLD LEG OF

/

AFFECTED UtlIT/

s ew Q

i

=w H

240

~

.a O

3 180 120 60 90025009 0-1 I

I I

1-0 200 400 6 0 0.. r.

800 1000 120'.

FlGURE B-7 l

540

'STEAf1 GEtlERATOR TEt1PERATURES t

VS l

tit!E j

480

\\

420 u.

O o

UtlAFFECTED UtlIT 360

~

wer

- H C"e wn.r 300 w

i-e AFFECTED UtlIT o

CCe 240 w

Z W

e a

E.

CC W

180 120 60 90025010-0 __

I I

I

_.L I

0 200 400 600 BCD 1000 120" TiliE ISEC) r

. = _..

1800 _

' FIGURE C-1 itAlti FEEDUATER FLO'.l V5 TIttE 1600 1400 uW to S

1200 to a

3o J

u.

1000 cr:

w F--

T l

2 1

OW 800 w

z C

r 600 i.

400 90025011 AFFECTED UilIT UNAFFECTED UilIT t

0 i

i I

I t

f 0

200 400 600 800

C00 120:

TINE (SECi l

r

.a i

f 360 FIGURE C-2 320 E'iCEGC"CY ICCD',.'ATER FLOll VS TIF.E 280

!!0 FL0ll TO u

UtlAFFECTED UllIT w

g g

r l.

m d

240 v

xo J

k gg w

200 w

C" I

2 OW W

u.

160 3u zW o

cc W

E 120 m

80 40 S6025012.

~

O l

I I

O 200 400 60C SCO 1000

2C, TIME (SEC)

FIGURE C-3 m

8 REACTIVITY CHAilGES VS TItiE i

6

!!ODERATOR 4

2 s

z w

a cc 4

W DOPPLER

~

0 3

i n

n u

BOR0tl C

-2 g

cc

~4 TOTAL

-6

)

i

-8 90025013

~

CEA

-10 I

i 1.

I I

0 200 400 -

600 800 1000 120".

TIME.fSEC) j 1

+

FIGURE C-4 0.90 CORE POWER vs.'

'TI!!E 0.80 0.70

~

0.60 0.5 0 e

W C

a a.

m M

0.t0 O

U 0.30 t

1 0.20

\\

l 0.10 0025014 0-1 1

1 1

0.0 t

0 200 400 600 800 1000 1200 TIME [SEC) l

F2GURE c-5 55 C0!!TAlt!!!EllT PRESSURE

.vs TI!!E 50 4s

.a E

40 t

.W CC "D

4 me m

as 1

cc L

-z m

zz 30

-e p

2 O

U 2s 20 I

15

~

l cQ025015 10 I

i i

I L

0 200 400 -

600 800 1000 120 TIME (SEC1 e-----

FIGURE C-6 PRI!1AR LOOP TDiPERATURES 540 y3 J

tit 1E 480 420 COLD LEG OF UNAFFECTED UtlI,T

~

u.

HOT L.EG 360 ow O

w

&a 300 COLD LEG OF cc AFFECTED UNIT /

W 3.

2-W i

240 o_o C)

J-180

.l-120 60 98025016 O

~

I i

t t

0 200 400 6C0 800 1000 1200 TIME (SEC1

FIGURE C.7 540 STEN 1 GEflERATOR TEi1PERATURES VS.

TIf1E e

F 480 420 LL.

Ow o

UllAFFECTED UtlIT 360 to cco C

cc tu n.r 300 m

AFFECTED H

UNIT cc O

ct ce 240 tu z

w a

r C

180 v>

120 60 0025017 0

I I

I I

1 0

200 400 600 800 1000 120'.

TIME fSEC1

FIfiURE D-1 1800 i%Ifl FEEDilATER FLO!!

VS J

TIllE 1600 1400 uW m

N 1200 CO

.J 2o J

u.

1000 c,:

ww cc 2

Ow 800 wu.

z cc 600 400 l

.200 AFFECTED UilIT UllAFFECTED UtlIT e

.8 0

1 I

I I

O 200 400 600

~800 1000 120:

TIME (SED)

k i

t c

360 f

FIGURE D-2 320

~

Et1ERGEllCY FEEDilATER FLO!!

VS i

TIllE fl0 FLO!! TO 260 UtlAFFECTED Ul LIT u

a y

v3 s

r j

co d

240 $

1o cc w

200 n

c 1

O w

w g

ISO 3u Z

i w

o cew 120 4

)

60 40 90025019 I

O I

I t

1 0

200 400 600 800 1000 120.

TIME (SEC)

I 1(iUl:L D-3 RCACTIVITY CilAllGES VS tit 1E B

g e.

6 t100EPATOR

,s 4

2 r

zwu DOPPLER 0

I H

~

n U

c

-2 TOTAL wx TOTAL

_4 BOR0il i

-6 l

CEA

-B w

90025020

-10 I

i t.

1 1

'O 200 400 600 800 1000 1200 TIME (SEC)

a s

g.>

0.90 FIGUf}E D-4 CORE P0llEP.

I vs 0.80 TIttE l

t.

i s

I 0.70 O.60 0.50 x

w

=c a

a.

w e

0.40 0

U 0.30 l

t i

l 0.20 0.10 i

90025021 I

I I

0.00 0

200 400 600 800 1000 1200 TIME FSEC) i

.. ;.i,-

FIGURE D-5 i

CONTA!!it1EllT PRESSURE 55 vs tit 1E 50 45 c

va 40 a

w tu cco Ln v>u 35 a:

a.

n Z

LU rz 30 c

s z

O U

25 s

t i.

20 i

15

~

90025022 10 I

I I

t._

f

'O 200 400 600 BCC 1000 1200 i

TIME (Sf.C1

... -.... w =.:.:; --= ~-.,

a..au.:.:

t FIGURE D-G.

540 T

'PRIliARY LOOP TE!1PERATURES vs i TIliE l

480 COLD LEG OF UNAFFECTED UNIT I

420 u.

.! LOT _l.ES, 360 a

uo

\\

(f) 1 w

cc COLD LEG OF o

300 F--

AFFECTED UNIT '

c-4 tr:

w Q_

s-w t.

240 a.o o

J 180 t

120 60 960250.23 0

t I

i i

i 0

200 400 600 800 1000 1200 TIME (SEC)

FIGURED-7 540 _

s'TEAfi GEllERATOR TE!!PERATURES vs TIttE

(

480 _

420 L

U!!AFFECTED UtlIT a

w 360 w

e s

C e

w Q.

300 g

AFFECTED H

UtlIT eo r-C M

240 W

0 o.

r C

W 180 W

e.

120 60 90025024 O

I I

I I

I O

200 400 600 800 1000 12C:

TIl1E ISEC)

' I; Enclosure (2) b to i

CE-18074-706 RETURN TO POWER ANALYSIS -

This attachment illustrates the impact of automatic initiation of auxiliary feed-water system (AFWS) on the licensing cases analyzed for a Main Steam Line Break Event. The provided results of analysis for the most limiting Main Steam Line Breaks (MSLB) with respect to return to power assume this system is designed with a delay in the automatic delivery'of the auxiliary feedwater to the steam genera-tors. The analyses performed for this evaluation, therefore, allowed for'a three (3) minute delay in the actuation of the AFWS subsequent to receiving a low steam generator water level trip signal. The conclusions of this analysis are considered to be applicable to Fort Calhoun.

The most limiting cases analyzed were those select'ed from the Design Basis Events analyzed in the FSAR or subsequent reload licensing submittals, as appropriate.

The most limiting case was found to be a full power-full flow Main Steam Line Break inside containment with automatic initiation of AFWS after a 3 minute delay.

The analysis was continued until the subcriticality margin was continuously in-creasing. The delay of 3 minutes assumed as part of the design of the automati-cally initiated auxiliary feed system was modeled conservatively in the analysis.

The MSLB outside containment is less limiting because the blowdown rate of the steam generators is restricted by the flow venturies located in the~ steam lines thus leading to a less severe reactivity insertion and a smaller potential for return-to-power than the results presented herein.

The results of the limiting case show that the affected steam generator blows dry in about 70 seconds and begins Reaci.cr Coolant System (RCS) cooldown with feed-water only. The peak power level attained including decay heat and subtritical multiplication is 12%.

From the time the steam generator runs dry until the actuation of auxiliary feedwater system, boron injected by High Pressure Safety Injection (HPSI), actuated at about 16 seconds into the transient, continues to add more negative reactivity to the core. After the initiation of AFW flow, the cooldown of the reactor coolant system (RCS) is resumed. The auxiliary flow is conservatively assumed to feed the affected steam generator only. The assumed auxiliary flow to the affected steam generator was conservatively taken to be about 20% of the full power feedwater flow. The continued cooldown of the RCS adds more positive reactivity which is eventually terminated oy the Low Pressure Safety Injection (LPSI) flow injected due to low RCS pressure. The negative reactivity inserted via LPSI flow terminates the reactivity excursion. The return-to-power attained after the AFWS delivery is 10.7%. Thus, with the 3 minute delay in the actuation of AFWS, the auxiliary feedw3ter will be introduced away from the most critical time frame with respect to return-to-power and the con-clusions of the MSLB events presented in the FSAR and the subsequent reload s'ubmittals conservatively bound those with the controI grade automatic initia-tion of auxiliary feedwater systems included. A typical sequence of events, typical for operating C-E plants, for the limiting case are presented in Table 1.

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l The MSLB results presented in the FSAR and subsequent reload licensing submittals l

assumed the following consequential failures in addition to the single failure which initiates the event (i.e., the double ended pipe break'inside containment):

(a) On reactor scram, the highest worth Control Element Assembly.

is assumed to stick in the fully withdrawn position, 1

(b) On Safety Injection Actuation, on the HPSI and one of the LPSI safety injection pumps are assumed to fail to start.

(c)NomainfeedwaterisolationisassumedonMSIS. The main feed flow is assumed to coastdown to 5% of full power flow in 60 seconds.

(More realistically flow would ramp to zero in about 20 seconds.)

Single failures were considered in the design basis to the extent that a failure initiates the event and safety grade equipment is designed to accommodate single failures as described above and is consistent with the design basis presented in the FSAR.

No consequential failures other than previously identified were con-

~

sidered. All control systems considered were assumed to function in the manner consistent with the FSAR.

Single failures concurrent with the MSLB (other than those identified above), as well as loss of offsite power concurrent with MSLB, are not, and have not been part of the design basis as described in the FSAR and, therefore, were not con-sidered.

l 90025026

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N

. TABLE A-1 Sequence of Events for the Main Steam Line Break Event with Automatic Initiation of Auxiliary Feedwater System (Full load, Two-Loop Condition Nozzle Break) d-Time (sec.)'

Event SafetySysteNInitiated Setpoint or Value

'.0.0 Initiation of break'

)

4, 3.4 Low steam Generator Pressure Reactor Prot'ection System

~478 psia trip signal occurs, f1 SIS Main Steam Isolation System initiated and liain Steam Isolation Valves begin 4

to close.

';i

'k:

A'

'4. 3 Trip breakers open q

y 4.8 CEAs begin to drop into Reactor Protection System Core 10.7 Comple.te closure of Main Steam 1 solation Valves to terminating blowdown from the intact steam generator 15.9 Pressurizer empties 16.2 Low RCS pressure, SIAS Safety Injection System 1563 psia Initiated 22.8 High Pressure Safety Safety Injection System 1220 psia Injection flow Initiated I

64.8 Main feedwater flow completes ramp down to 5%

68.7 Affected steam generator liquid inventory depleted and beginning of blowdown of

,.2,-

l feedwater only

Q. y

• '.G.?..

71.9 Peak return-to-power

• occurs i

12%

w an a peak reactivity f.

of.186%op 4

A i

return-to-power includes decay heat and suberitical multiplication 90025027 i

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-TABLE.'A-1 ( y tinued) r i

i Time.(sec.)

Event E

Safety System Initiated _

Setpoint or Value 150.0

. Auxiliary Feedwater flow to:

Auxiliary Feedwater System

'affected steam generator initiated j

31 8.7 Low Pressure Safety Injecti3n Safety Injection System i;!

207 psia flow Initiated

.3 319.9 Peak reactivity post i,----

auxiliary feedwater

+.13%Ap delivery 345.1 Peak return to power post auxiliary feedwater delivery 10.7%

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