Text

Updated Final Safety Analysis Report (Ufsar), Revision 32, Chapter 14, Section 14.1, Figures 14.1.0-1 to 14.1.12-2 (Unit 2)
	ML24159A179
Person / Time
Site:	Cook
Issue date:	05/30/2024
From:	; Indiana Michigan Power Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML24159A261	List: AEP-NRC-2024-20, Updated Final Safety Analysis Report (Ufsar), Revision 32, Chapter 3, Reactor (Unit 2); ML24159A163; ML24159A164; ML24159A165; ML24159A166; ML24159A167; ML24159A168; ML24159A169; ML24159A170; ML24159A171; ML24159A172; ML24159A173; ML24159A174; ML24159A175; ML24159A176; ML24159A177; ML24159A178; ML24159A179; ML24159A180; ML24159A181; ML24159A182; ML24159A183; ML24159A184; ML24159A185; ML24159A186; ... further results
References
	AEP-NRC-2024-20
	Download: ML24159A179 (1)
	v • d • e

Revision: 19.1 Change

Description:

UCR-1719

Title:

Doppler Power Coefficient Used In Safety Analyses AMERICAN ELECTRIC POWER (where 100% power is 3588 MWt)

COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.0-1 Sheet 1 of 1

UNIT 2

s 0.3 N

3 0.2

f 2 0.1

RCCA POSITION (FRACTION VS. NOMINAL) 1

Figure 14.1.0-3 NORMALIZED RCCA REACTIVITY WORTH VS. RCCA POSITION

UNIT 2 July, 1993 1

0.9

0.8

0.7

0.8

0.1

0.4 3 0.3 N -

0.2

0.1

0 0 1 2 3

TIME AFTER RCCA DROP BEGINS (SECONDS)

Figure 14.1.0-4 NORMALIZED RCCA REACTIVITY WORTH VS. TIME AFTER RCCA DROP BEGINS I

UNIT 2 July, 1993 2400 PSIA

\\ \\ \\ \\ \\ \\

\\ \\ \\ \\ \\ \\ \\

\\ \\ \\ \\ \\

\\ \\ 8 \\ * \\ \\

45 STEAM GENERATOR 'J \\ \\ \\ \\ \\ _. \\ \\

SAFETY VALVES OPEN

40 S60 565 570 S7S 580 58s 5% 595 699 695 610 615 626 62S

OTAT Protect ion Lines

Core Thermal Safety Limits

Figure 14.1.0-s Overtemperature and Overpower AT Protection

Core Conditions

- Transition Cycles

- Nominal Vessel Average Temperature = 576'F

- Nominal Pressurizer Pressure - 2250 psia

UNIT 2 July, 1993 I

65 ? I \\ \\ \\ \\ \\

\\ \\ \\ \\ / \\ \\ \\ \\ \\

\\

PaA \\ \\ \\ 1 \\

\\ \\

\\ \\ \\ \\ \\ \\ \\ \\ \\ \\

I/\\, \\ \\ \\ \\ \\ \\ LA 2250 \\ PSIA \\

SAFETY VALVES OPEN

'5 589 SOS 690 695 619 615 629 625 654 8

OTAT Protection Lines

Core Thermal Safety Limits

Figure 14.1.0-6 Overtemperature and Overpower AT Protection

Core Condi'tions:

- Full VANTAGE 5 Core

- Nominal Vessel Average Temperature = 581.3.F

- Nominal Pressurizer Pressure - 2100 psia.

UNIT 2 July, 1993 I Revision: 18.1 Change

Description:

UCR-1630

Title:

Revised Overtemperature and Overpower T Protection AMERICAN ELECTRIC POWER Core Conditions: - Full VANTAGE 5 Fuel COOK NUCLEAR PLANT -Nominal Vessel Average Temperature = 581.3°F NUCLEAR GENERATION GROUP -Nominalressuzer Pssu100sia BRIDGMAN, MICHIGAN (section4.1.0.6-1or discsio UFSAR Figure: 14.1.0-6A Sheet 1 of 1

UNIT 2 101

100 101 102 TIME AFTER SHUTDOWN (SECONDS)

figure 14.1.0-7 1979 ANS Residual Decay Heat Used In Accident Analyses

UNIT 2 July, 1993 Revision: 18 Change

Description:

UCR-1611

Title:

Rod Withdrawal from Subcritical Nuclear Power and Heat F lux AMERICAN ELECTRIC POWER Versus Time COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.1-1 Sheet 1 of 1

Unit 2 Revision: 18 Change Description : UCR-1611

Title:

Rod Withdrawal from Subcritical Fuel Average AMERICAN ELECTRIC POWER and Clad Temperatures Versus Time COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.1-2 Sheet 1 of 1

Unit 2

0. i 0 2 4.6 a 10 12 14 16 ;a 2 0 TIrnE ISECI

Figure 14.1.26-l Rod Withdrawal at Power Nuclear Power Versus Time for Full Power, 80 PCM/Sec Insertion Rate, Maximum Reactivity Feedback

UNIT 2 July 1991

600

560

0 2 4 b 8 10 12 :4 lb :3 20

4.

3.5

3.

2.S

2.

I.5

1. *, I 0 2 4 b. a 10 12 14 16 18 20

TIN ISECI

figure 14.1.28-3 Rod Withdrawal at Power Core Average Temperature and ONBR Versus Time for Full Power, 80 PCM/Sec Insertion Rate, Maximum Reactivity Feedback

UNIT 2 July 1991 0 2s so. 7S 100 12s IS0 17s 200 225 250 275 300 325 350 375 400 1

TIflE 1 SEC 1

figure 14.1.26-4 Rod Withdrawal at Power Nuclear Power Versus Time for Full Power, 4 PCM/Sec Insertion Rate, Maximum Reactivity Feedback

UiJIT 2 July 1991 1800. J /

0. so. 100. lS0. 200. 250. 300. 350. 400.

2000.

1800. l

600.

1400.

1200.

1000.

800. C

0. so. 100. 150. LOO. tso. 300. 350. 400.

TIM cSEc)

Figure 14.1.28-5 Rod Withdrawal at Power Pressurizer Pressure and Water Volume Versus Time for Full Power, 4 PCM/Sec Insertion Rate, Maximum Reactivity Feedback

UNIT 2 July 1991 660

640

600 --

0 2s so 75 100 125 IS0 17s 200 225 2;O 27s 300 325 350 375 400

4.

3.5

3.

2.s

2. J

I.5

1.

0 2s so 75 100 12s 150 175 200 22S 250 27s 300 325 350 37s 400 rIna (SEC)

Figure 14.1.28-6 Rod Withdrawal at Power Core Average Temperature and ONBR Versus Time for Full Power, 4 PCM/Sec Insertion Rate, Maximum Reactivity Feedback

UNIT 2 July 1991

--- -I_ -- --

.-- 7 ---. _ _

--.--- ------we_..

_._-.-- a--*.- - - --.. - - ---.

-..-.-.. a m-- % i.+-- : -.--..-a A - _. _._ _ __ __

ao zz 2 i

-- - X..- --- --_--- - - - --.

33 Y

x ! -.

1

-- -.-- \\ I \\ -- f \\

--. I \\ I.

\\,

-.-_ L

.-.: _ -....: 1 f :

-..--. \\ \\ --. --- a- --------. - \\

\\

-- ----.a--._-_--_--_

.-_ _. e---p-

--.- )--; -/---....- ^ ---_ -

- L.

-. :. : i.- :.. I

-_. ---_ 1 ----- - --.- -- _... -

L b - =- I*- :

t:z.: : / --

..-M--w.. I... --.

. ---- -. I _ - _

I : i - - --- ---

UNIT 2 July 1991 1.2000 L 1.2000 L

i i 8 8 - -

g ~80000 l g ~80000 l

i 70000 9. i 70000

t 60000 + 60000 t

Figure 14.1.3-1 Dropped RCCA(s)

Nuclear Power and Core Heat Flux Versus Time for a Typical Response in Automatic,Control

UNIT 2 July 1991

+

ii t 550.0 l 9

Y

9 530.0 l w

I Y

s l 510.0 l..

L.

1900.0 --

s.

i 1800.0 l.

1700.0 -9..

4 c Q 8 4 8 r! ?

0 d si s, e g ie s

nnt csatt

Figure 14.1.3-2 Dropped RCCA(s)

Average Coolant Temperature and Pressurizer Pressure Versus Time for a Typical Response in Automatic Control I

UNIT 2 July 1991 0 1 L 3 4 6 6 7 a 9 10 TIflE [SEC 1

figure 14.1.6-l Complete Loss of Flow Core Flow Coastdown Versus Time

UNIT 2 July 1991

.8

.b

.4

.2

0.

0 1 2 3 4 5 b 7 8 9 10

2640.

2500.

2400.

2300.

2200.

2100.

2000.

1900.

1800.

0. 1. 2. 8. 4. 6. b. 7. 8. 9. 10.

TlnL (SEC1

Figure 14.1.6-2 Complete Loss of Flow Nuclear Power and Pressurizer'Pressure Versus Time

UNIT 2 July 1991

2.0

2.6.

1.8

1.6

1.4 0 1 2 3 4 5 6 7 8 9 10 TIME (SEC)

figure 14.1.6-4 Complete Loss of Flow DNER Versus Time

UNIT 2 July 199i A.4

1.2

1.

.e

.6

.4

.2

0.

0 1 2 3 4 6 6 7 8 9 10

1.4

1.2

1.

.8

.6

-4

.2

0.

0 1 2 8 4 s 6 7 0 9 10

TIllE (SEC)

figure 14.1.6-5 Partial Loss of Flow l/4 Faulted Loop and Core Flows Versus Time

UNIT 2 July 1991 3

P

0. - 0 1 f 3 4 5 4 1 8 9 1 10

2bOO. -

2500.

2400.

2300.

2200.

2100. L

2000..

1900...

1.00. a

0. r 1. g. 3. 4. 5. 6. 7. a. 9. 10.

TIM fSCCl

Figure 14.1.6-6 Partial Loss of Flow l/4 Nuclear Power and Pressurizer Pressure Versus Time

UNIT 2 July 1991 1.4

1.2

I.

l 8

b

l 4

.2

0.

1.4

I.2

1.

.I

.b

l 4

.2

8. c

Figure 14.1.6-7 Partial Loss of Flow I/4 Average Channel and Hot Channel Heat Flux Versus Time

UNIT 2 July 1991 5.0

4.5

4.0

3.5 a

s a

3.0

2.5

2.0

1.5 \\

0 I 2 3 b 5 5 7 8 9 10

TIME (SEC)

Figure 14.1.6-8 Partial Loss of Flow l/4 DNBR Versus Time

UHIT 2 July 1991 Revision: 18.1 Change

Description:

UCR-1630

Title:

Total Core Flow and Faulted Loop Flow vs. Time For AMERICAN ELECTRIC POWER The Locked Rotor Event COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.6-9 Sheet 1 of 1

UNIT 2 Revision: 18.1 Change

Description:

UCR-1630

Title:

Nuclear Power and RCS Pressure vs. Time For The AMERICAN ELECTRIC POWER Locked Rotor Event COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.6-10 Sheet 1 of 1

UNIT 2

.. G -

. ~-----y

\\,

I

.3 -.. 5.

Figure 14.1.6-11 l/4 Locked Rotor Average Channel and Hot Channel Heat Flux Versus Time

UNIT 2 Julv 1991 ii?O. T /

I,.f

tsi. 7 j

300. -. i ii

Core Average and Loop 1 Temperature vs. Time For AMERICAN ELECTRIC POWER Loss of Load, Maximum Reactivity Feedback Without COOK NUCLEAR PLANT Pressurizer Spray and PORVs NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.8-12 Sheet 1 of 1

UNIT 2 I

' i

! I

.6 i

.A :

I

.2 j

I a.

.6

Figure 14.1.9-1 Loss of Normal Feedwater Nuclear Power and Core Heat Flux Versus Time

UNIT 2 July, 1992 Revision: 20.2 Change

Description:

UCR-1815

Title:

Loss of Normal Feedwater Loop Temperature AMERICAN ELECTRIC POWER Versus Time COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.9-2 Sheet 1 of 1

UNIT 2

20 40 60 80 100 120 140 160 180 i TIME (SEC)

540-p

520-u

=o 4 20 40 60 80 100 120 140 160 180 :

TIME (SEC)

DONALD C. COOK NUCLEAR PLANT UNIT ~(FULLVSCORE)

NUCLEAR POWER TFMNSIENT Kc CORE AVERAGE TEMPERATURE vs. TLME FOR THE SINOLE LOOP FEEDWATER MALFUNCIION WlTH AUTOMATIC ROD CONTROL AT FULL POWER L Figure 14.l.lOB-1

UNIT 2 July, 1993

2700.

z ;; 2600.

n

- 2500.

2 g 2400. t z

n 2300.

I2 z 2200.

iz I z 2100.'

E I -

2000.

1900.

I 1800.1 0. 20. 40. 60. 80. 100. 120. 140. 160. 180. 21 TIME (SEC)

5.

4.5

4.

3.5 fs B 3.

2.5, a-

2.

1.5

1. c 0 zo 40 60 80 100 120 140 180 160 i

TIME (SEC)

DONALDC.COOK NUCLEARPLANT

UNIT 2(FULL vs CORE)

PRESSURIZER PRESSURE k DNBR 8. TIME FOR THE SINGLE LOOP FEEDWATER MALFUNCTION WlTH AUTOMATIC ROD CONTROL AT FULL POWER

Figure 14.1. IOB-2

UNIT 2 July, 1993 680

680 G:, 040

3 620

a 000 h e

520 t

500 0 20 40 60 80 100 120 140 160 160 2 TIME (SEC)

DONALD C. COOK NUCLEAR PLANT

UNIT 2(FULL v5 CORE)

NUCLEAR POWER ?RANSIENT & CORE AVERAGE TEMPERATURE va. TIME FOR THE SINGLE LOOP FEEDWATER MALFUNCTION WITH MANUAL ROD CONTROL AT FULL POWER Figure 14.l.lOB-3

UNIT 2 July, 1993 2 ;; 2600.

h t 2500.-

2400. --

2300.'.

2200. *'

21oo.q

2000.-

lQOO.*-

1800. l 0. 20. 40. 60. 80. 100. 120. 140. 160. 180. 2 TIYE (SEC)

5.

1.4 0 20 40 60 80 100 120 140 180 180 2 TIME (SW

DONALD C. COOK NUCLEAR PLANT UNIT ~(FULLVSCORE)

PRESSURIZER PRESSURE & DNBR VI. TIME FOR THE SINGLE LOOP FEEDWATER MALFUNCTION WITH MANUAL ROD CONTROL AT FULL WWER

Figure 14.l.lOB-4

UNIT 2 July, 1993 1.4

1.2

1.

.a

.6

.4

.t

0.

0 20 40 60 80 100 120 140 160

700

660

640

620

600

580

560

540

520

500 u 60 10 100 120 140 160 110 200 TIRE (SEC)

DONALD C. COOK NUCLEAR PLANT UN-IT ~(FULL vs CORE)

NUCLEAR POWER TRANSIENT & CORE AVERAGE TEMPERATURE VI. TIME FOR THE MULTI-LOOP FEEDWATEX MALFUNCTION WITH AUTOMATIC ROD CONTROL AT FULL POWER

, Figure 14.1.10B-5

UNIT 2 July, 1993 2700.

2600.

2500.

2400.

2300.

2200.

2100.

2000.

1900.

1100,

1700.

20. 40. 60. 10. 100. 120. 140. 160. 110. 200.

TIWE (SEC)

5.

4.5

4.

3.5

2.5

0 20 40 60 10 100 120 140 160 110 200 TIM (SEC)

DONALDC.COOK NUCLEARPLANT UNIT ~~ULLVXORE)

PRESSURIZER PRESSURE & DNBR VL. TIME FOR THE MULTI-LOOP FEEDWATER MALFUNCI-ION W-ITH AUTOMATIC ROD CONTROL AT FULL FOWER

Figure 14.l.lOB-6

UNIT 2 July, 1993

.b

.c

.2

0. 0 20 40 60 10 100 120 140 160

TIHE :'SEC)

700

610

660

640

620

600

510

560

540

520

500 -- -

0 LO 40 60 10 100 120 110 160 110 200 TIHE (SEC)

DONALD C. COOK NUCLEAR PLANT UNIT ~(FULLVSCORE)

NUCLEARPOWERTRANSIENT&CORE AVERAGETEh4PERATUREv~.TIMEFOR THEMULTI-LOOPFEEDWATER MALFUNCTIONWITHMANUALROD CONTROLATFULLPOWER Figure 14.1. IOB-7

UNIT 2 July, 1993 1

2700.

2600.

zsoo.

2400.

2300.

2200.

2100.

2000.

1900.

1100.

1700.

0. 20. 40. 60. 80. 100. 120. 140. 160. 1ao. 200.

TIRE (SEC)

5.

4.5

4.

3.5

0 20 40 60 10 100 120 140 160 110 200 TIRE (SEC)

DONALD C. COOK NUCLEAR PLANT UNIT SKULL vs CORE)

PRESSURIZER PRESSURE & DNBR vs. TIME FOR THE MULTI-LOOP FEEDWATER MALFUNCI7ON WTH MANUAL ROD CONTROL AT FULL POWER Figure 14.1. IOB-8

UNIT 2 July, 1993 0 20 40 bO 80 100 120 140 1 CO 180 200 226 240 2bO 280 300

azso..

2000. + i A-

I 6 17fO.

1500. I

12SO

1000. !

750..

500. *-

aso. *-

0. t !

0. 40. 80. 120. 1 co. LOO. 240. 280.

TlnE (SEC1

Figure 14.1.118-l Excessive Load Increase Nuclear Power and Pressurizer Pressure Versus Time for Minimum Reactivity Feedback w ith Manua 1 Rod Control

UNIT 2 July 1991 so0 0 20 40 co a0 100 :20 I40 I bO 180. 200 220 240 zb0 290 300

3.2.

3. a.

- - i i s.;

0 - 2.b

i 2.4

5 0 2.2

0 u

2.

3 u d 1.8

0 l.b

1.4.

1.2

I

1. C

0 20 40 co 80 100 120.I40 I co 180 200 220 240 Lb0 280 300

rlnc I SEC 1

Figure 14.1.118-2 Excessive Load Increase Core Average Temperature and ONBR Versus Time for Minimum Reactivity Feedback with Manual Rod Control

UNIT 2 July 1991 I. 4 F r

.. t - -

J.

0 20 40 bo 80 100 120 I40 I CO 180 200 220 240 2bO 2ao 300

tsoo.

czso. *-

3000. I

1750.

1 1500. --

LZSO...

1000..-

7s 0.. *-

SOO. *-

214-a *-

0. J 4

0. 40. 80. 120. 1 b0. 200. 240. 280.

TIRE (SEC1

Figure 14.1.118-3 Excessive Load Increase Nuclear Power and Pressurizer Versus Time for Maximum Reactivity feedback with Manual Control

UNIT 2 July 1991 0 20 40 LO 80 100 120 140 lb0 I80 200 220 240 260 280 300

1. J 0 20 40 LO 80 100 120 140 lb. 18@ 200 220 240 2bO 280 300

ttnc ISlCl

Figure 14.1.118-4 Excessive Load Increase Core Average Temperature and ONBR Versus Time for Maximum Reactivity Feedback with Manual Control I

UNIT 2 July 1991 2000. +

IfSO..-

LSOO. --

1250..-

1000..-

tso...

500..-

tso. *-

0. I

0. 40. 80. 120. 1bO. 200. 240. 280.

trn2 I stc I

Figure 14.1.11B-5 Excessive Load Increase Nuclear Power and Pressurizer Pressure Versus Time for Minimum Reactivity Feedback with Automatic Rod Control

UNIT 2 July 1991 se0

LfO

IL0

550

s40

530

520

510

500 0 20 40 co 80 100 f20 140 lb0 180 200 22c 240 260 280 300

2.L

2.4

2.2

2.

1.8

1. L

1.4

1.2

1.

0 20 48 co 80 100 120 148 lb0 18k 108 221 24I 2bO 2so 30) rlnc I see I

14.1.118-6 Excessive Load Increase Core Average Temperature and DNBR Versus Time for Minimum Reactivity Feedback with Automatic Rod Control I

UNIT 2 July 1991

.b

.4

. 2

a.

0 20 40 b0 80 100 I20 I40 lb0 180 200 22'J 240 260 280 300

!OOO*

17SO.

1500.

izso.

1000.

7so.

soo.

aso.

0.

0. 40. 00. 120. 1 b0. 200. 240. 2IO.

TIRC I SEC I

Figure 14.1.116-7 Excessive Load Increase Nuclear Power and Pressurizer Pressure Versus Time for Maximum Reactivity Feedback with Automatic Rod Control

UNIT 2 July 1991 sea c

570.

SbO *-

550 --

s40.-

s30.

520

510 *-

SO0 i 0 20 40 LO 80 100 I20 140 I60 180 200 220 240 are 280 300

3.2 y

3. *-

3.3 i i

2.b.'

2.4 --

0 20 40 b0 a0 IO0.t@ 140 160 1.0 200 220 240 2bO 200 300 tint 1 stc I

Figure 14.1.118-8 Excessive Load Increase Core Average Temperature and ONBR Versus Time for Maximum Reactivity feedback with Automatic Rod Control

UNIT 2 July 1991 Revision: 20.2 Change

Description:

UCR-1815

Title:

Loss of Offsite Power to the Station Auxiliaries AMERICAN ELECTRIC POWER Nuclear Power and Core Flow Versus Time COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.1.12-1 Sheet 1 of 1

UNIT 2 rzo. ;

@DO. : I I THOT

Figure 14.1.12-2 Loss of Offsite Power to the Station Auxiliaries Loop Temperature and Pressurizer Water Volume Versus Time

UNIT 2 July, 1992

ML24159A179

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