Text

Rev. 29 to Updated Safety Analysis Report, Section Xiv, Figures
	ML19137A150
Person / Time
Site:	Cooper
Issue date:	04/24/2019
From:	; Nebraska Public Power District (NPPD)
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML19137A206	List: ML19137A122; ML19137A123; ML19137A124; ML19137A125; ML19137A126; ML19137A127; ML19137A129; ML19137A130; ML19137A131; ML19137A132; ML19137A133; ML19137A134; ML19137A135; ML19137A136; ML19137A137; ML19137A138; ML19137A139; ML19137A140; ML19137A141; ML19137A142; ML19137A143; ML19137A144; ML19137A145; ML19137A146; ML19137A147; ML19137A148; ML19137A149; ML19137A150; ML19137A153; ML19137A155; ML19137A156; ML19137A157; ML19137A158; ML19137A159; ML19137A160; ML19137A161; ML19137A162; ML19137A163; ML19137A164; ML19137A165; ML19137A166; ML19137A167; ML19137A203; NLS2019023, Rev. 29 to Updated Safety Analysis Report, Enclosure 4B, Reference Drawings;
References
	NLS2019023
	Download: ML19137A150 (98)
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CADD Fl LE: C0048885 NUCLEAR SYSTEM PARAMETER VARIATION EVENT OPEN/CLOSE ANY VALVE STOP/START ANY COMPONENT ELECTRICAL FAILURE MANIPULATE ANY CONTROL DEVICE SINGLE OPERATOR ERROR TYPES EVENT APPLJCATJON EVALUATION OF DAMAGE OF RADIOACTIVE MATERIAL BARRIERS FUEL BARRIER REACTOR COOLANT PRESSURE BOUNDARY Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

STATION SAFETY ANALYSIS-METHOD FOR IDENTIFYING AND EVALUATING ABNORMAL OPERATIONAL TRANSIENTS FIGURE XIV-4-1 08/01/03

CATEGORY OF RADIOACTIVE MATERIAL RELEASE DD FILE: C0048886 ACCIDENT TYPE COMPONENT MECHANICAL FAlLURE FUEL OVER-HEATING PIPE BREAK COMPONENT MECHANICAL FAILURE FUEL OVER-HEATING PIPE BREAK COMPONENT MECHANICAL FAILURE FUEL OVER-HEATING PIPE BREAK COMPONENT MECHANICAL FAILURE FUEL OVER-HEATING PIPE BREAK COMPONENT MECHANICAL FAILURE FUEL OVER-HEATING ACCIDENT APPLICATION EVALUATION OF DAMAGE TO RADIOACTIVE MATERlAL BARRIERS SEC-ONDARY RELEASE OF RADJOACTIVE MATERIAL RADIOLOGICAL EFFECTS OFFSITE AND ONSITE DOSE OFFSITE AND ONSITE DOSE OFFSITE AND ONSITE DOSE i2i~= 1---------t-"'1 OFFSITE AND ONSJTE DOSE MENT OFFSJTE AND ONSITE DOSE Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

STATION SAFETY ANALYSIS-METHOD FOR IDENTIFYING AND EVALUATING ACCIDENTS FIGURE XIV-4-2 08/01/03

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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR) 6 Generator Trip (Load Rejection) without Bypass BOC to MOCl: MELLLA-HBB Figure XIV-5-2a 01/21/19

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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR) 6 6

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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR) 6 Generator Trip (Load Rejection) without Bypass MOC2 to EOC : MELLLA-HBB Figure XIV-5-2c 01/21/19

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-UPDATED SAFETY ANALYSIS REPORT (USAR)
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Generator Trip (Load Rejection) without Bypass MOCl to MOC2 : ICF-HBB Figure XIV-5-2e 01/21/19
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Generator Trip (Load Rejection) without Bypass MOC2 to EOC : ICF-HBB Figure XIV-5-2f 01/21/19
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Generator Trip (Load Rejection) without Bypass EOC : MELLLA-UB Figure XIV-5-2g 01/21/19
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR) 0 6
Generator Trip (Load Rejection) without Bypass EOC: ICF-UB Figure XIV-5-2h 01/21/19 400 36C 320 280 E,
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR)
Turbine Trip with Bypass Figure XIV-5-3 07/22/96
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mo 4.0 4.0 evel (righ. ax,s)
-a-loid eacivity
~
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-?t-Doppjer Reactivity
-al;:-
Turbine Steam Flow 3.5
-al;:- Scram Reactivity 3.5 200
-+-* Feedvvater Flmv 80
-¢- Total Reactivity 3.0 3.0 60 f 2.5 2.5 150 w
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\\
-100
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-2.0
-2.0 0
2 3
4 5
6 0
2 3
4 5
6 Time (sec)
Time (sec)
Nebraska Public Power District COOPER NUCLEAR ST A TION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass BOC to MOCl: MELLLA-HBB Figure XIV-5-4a 01/21/19
"O
<I)
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2 3
4 Time (sec) 6 625 125 r "
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eve! {right axls)
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Vessel Steam Fio1,1,t Turbine Stearn Flow
~
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5 6
Time (sec)
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~
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-D.5
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Do e ress ise (right axis) 400 Safety Valve Flow Relief Valve Flow Bypass Valve Flow 2
3 4
Time (sec)
-e-
/oid eac ivity
--:),!:-- Doppler Reactivity
-.- Scram Reactiviiy
-..- Total Reactivity
-~
2 3
4 5
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass MOCl to MOC2 : MELLLA-HBB Figure XIV-5-4b 01/21/19 a
6 360 320 280 120 80 40 0
4.0 3.5 3.0 2.5 2.0 e C
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-j,(;-
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3 4
5 Time {sec) 625 G
500 -"O
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ii:
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'5
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-+-
Feedwaier Fla,v 80 150 60 E
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2 3
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6 Time {sec) 100 90 80 70 60
-0 Q) 1ii 50 0::
0 40 3!)
20 10 0
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E 1.0 0
0
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0
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-+-
Bypass Valve Fiow 2
3 4
5 Time (sec}
-'6-
/oid,eacdvity
-3/4-Doppler Reactivity
~
Scram Reactivity
-+- Tota! Reactivity
\\
\\\\
2 3
4 5
Time (sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass MOC2 to EOC : MELLLA-HBB Figure XIV-5-4c 01/21/19 t l s
6 360 320 280 iii S:
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3 4
6 Time {sec)
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-1&- Turbine Steam FlrnN
~
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0
-2C
...J
-100 -l-'t-+--+->-t-_,__>-+-+-t-+-+-+-t--1-+-+-+--"4-+-+-+-+-t-+-+--*-+ -40 0
2 3
4 5
6 Time (sec)
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~
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-ill;:-
Relief Valve F!ow
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5 Time {sec) 6 360 320 280
e 240 "' "'
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3 4
5 Time(sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass BOC to MOCl: ICF-HBB Figure XIV-5-4d 01/21/19
-0.5
-i.O
-1.5
-150
-a-leutro
!ux {ri hI axis 750 100 400
-a-D01'ne ress ise (ri ht axis)
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-,lk-Core Inlet Flow 90
-r-Relief Valve Flow 360
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Time (sec) 250 evel (ligtn axis,t 100 40 4.G
-a-loid eacrivity
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--s-Scram Reactivity 3.5 200
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4 5
6 0
2 3
4 5
6 Time (sec)
Time {sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass MOCl to MOC2 : ICF-HBB Figure XIV-5-4e 01/21/19
150 750 100 400
-e-
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--:ill,-
360
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I 20 8C I
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125
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3 4
5 6
0 2
3 4
5 6
Time {sec}
Time (sec) 250 100 4.0 4.!J eve! {right axis)
-e-Joid,ea ivity
~
Vessel Steam Flow
~
Doppler Reaciivrty
-.Ir-Turbine Steam Flow 3.5
-;k-Scram Reactivity 3.5 200
~
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2.5 2.5 150
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w
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-1.5
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-40
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-2.0 0
2 3
4 5
6 0
2 3
4 6
Time (sec}
Time (secj Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass MOC2 to EOC: ICF-HBB Figure XIV-5-4f 01/21/19
-0.,
1'l a::
150 ~-~-~-~-~-~~~-~--~~~~~
750 125 100 75 50 25 D
250 200 150 50 0
-50
-jk-Core Inlet Flow 625 125 0
2 3
4 5
Time {sec)
~-~-~-~-~-~-~~~~~~~-~
100 eve!{right axis)
~
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3 4
5 Time {sec}
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-0.,
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R 0
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C:.,
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0 0
0 ll<S.,
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15 1.0 0.5 G.O
-D.5
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4 5
Time (sec}
ht axis) 400 360 320 28C E
200 :,
V, V,.,...
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0 0
~-~-~-~-~-~-~~~~~~~~~
4.0
-a-aid eactivity 0
4't-Doppler Reactivity
-r-Scram Reactivity Total Reactivity 2
3 4
5 Time (sec}
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass EOC: MELLLA-UB Figure XIV-5-4g 01/21/19 t ::
-0.5
-10
-1.5
150.--------.-....-,--....-,---,----,*=c,--.-,-;-,--,,--,,----,r?',C'-:---;'"':":T=---., 750 125 100 50 25 0
250 200 150 100 50
-50
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-J,:-
0 2
3 4
Time (sec}
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rti ll:'.
~
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-Jir-Turbine Steam flow
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3 4
Time {sec) 6 30 60 40 20
()
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i:
0
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-2,0
~-----.--e---~D~o-e-P-re_s_s~'~is-e~(r-ig~l1_!_a_x~is-)~ 400
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-.1,,--
Relief Valve Flow
~
Bypass Valve Flow 2
3 4
Time (sec) 5 6
360 320 280 w
s 240 w
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200 ::i:
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E 0
0
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2 3
4 5
Time {sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Turbine Trip without Bypass EOC: ICF-UB Figure XIV-5-4h 01/21/19 6
d
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USARl Closure of One Main Steam Isolation Valve {MSIV)
Figure XIV-5-5 07/22/96
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT rusAR Closure of All Main Steam Isolation Valves (MSIVs)
Figure XIV-5-6 07/22/96
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USARl DEH Pressure Controller Output Fails High Figure XIV-5-8 07/22/96
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR' Inadvertent Opening of a Safety/Relief Valve Figure XIV-5-9 07/22/96
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c t-Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR)
Loss of Feedwater Flow Figure XIV-S-10 07/22/96
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c I-iii Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Auxiliary Power (Trip without Transfer)
Figure XIV-5-11 07/22/96
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Loss of Auxiliary Power (Loss of Grid Connection}
Figure XIV-5-12 07 /22/96
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I-al Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT rusAR:
Recirculation Flow Controller Failure - Decreasing Flow Figure XIV-5-13 07/22/96
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I-Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR' Trip of One Recirculation Pump Figure XIV-5-14 07/22/96
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0 (031~8 dO 1N3J 83dl Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR' Trip of Two Recirculation Pumps Figure XIV-5-15 07/22/96
1.
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SLO Pump Seizure for hp1 CYCLE 28 POWER: 68.5% RATED FLOW: 57.1% RATED 1 oo.------------------.--~-1e_L_1tr_o_n~F-lu-x----~
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PIO: 31070 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Single Loop Operation Pump Seizure Figure XIV-5-15a 02/22/17
-N*III'...
0
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT msAR)
Recirculation Flow Controller Failure - Increasing Flow Figure XIV-5-16 07/22/96
ff-B f-H
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT msAR)
Startup of Idle Recirculation Pump Figure XIV-5-17 07/22/96
400
-a-euiron lux (rig,taxis) 80D 100
-a-Dome ress ise tright axis) 400
~
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand BOC to MOCI : MELLLA-HBB Figure XIV-5-18a 01/21/19
400
-a-eutron lux rig l axis) 800 100 400
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand BOC to MOCl : MELLLA-HBB - 1 TBVOOS Figure XIV-5-18b 01/21/19
400 800 100 400
-a-eutron lux _rig taxis)
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand MOCl to MOC2 : MELLLA-HBB Figure XIV-5-18c 01/21/19
400
-a-Jeutron lux,rig, taxis) 800 100 400
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Nebraska Public Power District COOPER NUCLEAR ST A TION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand MOC2 to EOC: MELLLA-HBB Figure XIV-5-18d 01/21/19
400 800 100 400
-a-eutron iux, rig t axis)
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand MOCl to EOC: MELLLA-HBB-1 TBVOOS Figure XIV-5-18e 01/21/19
400
-e-eutron lux rig taxis) 800 100 Dome Press Rise (right axis) 400
-a-
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~-
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-+--
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand BOC to MOCI : ICF-HBB Figure XIV-5-18f 01/21/19
400 800 100 400
-a-eutron lux rignt axis) Dome ress ise _right axis)
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand BOC to MOCI : ICF-HBB - 1 TBVOOS Figure XIV-5-18g 01/21/19
400 800 100 400
-s-euiron lux rig t axis)
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-x-Ave Surface Heat Flux
~
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,r
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand MOCl to MOC2 : ICF-HBB Figure XIV-5-18h 01/21/19
400 ----------~~-~-~~-~. 300
-s-eutron lux rig 1t axis) 350 300 0
5
~
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-fill,:-
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--<1,- Core lniet Subcooling 10 Time {sec) 15 700 600 175 -----------,-~--~..,,..,--,....,------,- 105 Level (right axis)
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125 50 25 0
0 5
10 Time (sec)
\\
15 90 15 0
100 90 80 70 60 JD 2D 10
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5
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~
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I I I I I
15
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ro 360 320 280 240 G>
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~
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-2. 0 +-+-..._+-+-+-+-+-->--+-+--1--1--1--1-+-'+-+-+-+ -2. 0 0
5 10 Time(sec}
15 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand MOC2 to EOC : ICF-HBB Figure XIV-5-18i 01/21/19
400 f 800 10G 400
-s-Neutron Flux rig taxis)
-s-Dome ress ise right axis)
-x-Ave Surface Heal Flux
-x-Safety Valve Flow
~
Core inlet Flow 90 Rel!ef Valve Fiow 360 350
~
Core Inlet Subcooling 700
--+-
Bypass Valve Fiow 80 320 300 600 70 280
'ti' 250 500 2
--.... ~,,..,.....
e 60 2:40 a, n::
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand MOCl to EOC : ICF-HBB - 1 TBVOOS Figure XIV-5-18j 01/21/19
400 Neutron Flux rig taxis) 800 1CO Dome ise _right axis) 400
-e-
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~
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand EOC : MELLLA-UB Figure XIV-5-lSk 01/21/19
400
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand EOC : MELLLA-UB - 1 TBVOOS Figure XIV-5-181 01/21/19
400
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand EOC: ICF-UB Figure XIV-5-18m 01/21/19
400 800 100 400
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Feedwater Controller Failure - Maximum demand EOC : ICF-UB - 1 TBVOOS Figure XIV-5-lSn 01/21/19
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0 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR' ATWS - MSIV Closure - with ARI Figure XIV-5-19 07/22/96
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CD Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR' ATWS -
MSIV Closure -
No ARI (First 100 Seconds)
Figure XIV-5-20 07/22/96
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR)
ATWS -
MSIV Closure -
No ARI (Long Term Response)
Figure XIV-5-21 07/22/96
250
-a-Neutron Flux:(% Rated}
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--A-Core Inlet Flow (% Rated~
225
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~ 125 100 75 50 25 0
0 5
10 15 20 25 30 35 40 45
j{i T111>1'(Sj 21)0
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\\ ~
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30 40 50 60 70 80 Time (s)
--e-Toiat P..eactrtir.~fSr, 20 25 30 35 40 45 50 Time(s)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
ATWS - MSIV Closure BOC, 76.8% Flow, 3 SRVOOS, MAX SR V Setpoint Figure XIV-5-21a 02/08/16
ISOO 1~00 140()
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()
1000 2000 3()00 40CO 5000 Time (aec)
/01}0 9{)1)0 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
ATWS - MSIV Closure BOC, 76.8% Flow, 3 SRVOOS, MAX SR V Setpoint Figure XIV-5-21b 02/08/16
~ -
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR' ATWS - Turbine Trip with Bypass
- with ARI Figure XIV-5-22 07/22/96
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I' Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
ATWS - Inadvertent Opening of a Relief Valve - with ARI Figure XIV-5-23 07/22i96
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1-
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(J)
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1.osi-:----:---1,-------+------+------+-----
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0.8 TIME <SECl 1.2 1.6*10' Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Transient Response ofIORV at MELLL and EOC Figure XIV-5-23a 09/19/00
1-w w
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1..
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7.
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<SECJ I.6*10' Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Transient Response ofIORV at MELLL and EOC Figure XIV-5-23b 09/19/00
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"::1' Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT fHSAR' ATWS - Pressure Regulator Failure open - with ARI Figure XIV-5-24 07/22/96
275 Neutron Flux/% Rated}

0-- Average Surf ace Heat Flux rtYv Rated)

~
Core Inlet Flow(% Rated) 250 -e-Core Inlet Subcooling (Etu/lbm}
225 200 175
.., 150 i
<? 125 100 75 50 25 0
200 175 150 125 100
~
-:;; 75 a:
50 25 0
-25 0
5 10 15 20 25 30 50 Time(sJ
-e-- le-*.:B fin.ma,,-e Sep Slrtrt;:*
~
vesserSteamRw,2,*i%,Ra.retJ';

is-Tuffifule Sti?:am Ff.il",;,li f% Raw;t."J,

-e-Fee@~Rmvn-::.F~;.
-50 +---+---+--+----+--lc---+----+---+--+---l 10 15 20 25 30 35 40 45 50 Time [s) oOO ~-~----.--.----,---,---,--B---,:c----,'l~e~ss~e~I P~, *...,,-, ~Ri...,,e-,(p"'scci,---,
450 -
400 350 300
" 250 2l "'
a:
"'200 150 1UD 50 0

<$- Safety vatve Flow (% Rated}

r
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\\
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~
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0 10 20 30 40 50 60 90 100 Tm>e(s)
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5 10 15 20 25
31) 35 40 45 50 Time{s)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
ATWS - Pressure Regulator Failure BOC, 76.8% Flow, 3 SRVOOS,
+70 PSI SRV Setpoint Figure XIV-5-24a 02/08/16
jbl)()
1Mi!i 1400
-.n5o l
1300
.. l 1250 \\
11/
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'.l anuo WJOO Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
ATWS - Pressure Regulator Failure BOC, 76.8% Flow, 3 SRVOOS,
+70 PSI SRV Setpoint Figure XIV-5-24b 02/08/16
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D D co Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR)
ATWS -
Loss of Normal Feedwater
- with ARI Figure XIV-5-25 07/22/96
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USAR' ATWS -
Loss of Normal AC Power
- with ARI Figure XIV-5-26 07/22/96
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io llu 1.2 J.6*10' Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

Transient Response ofLOAP at MELLL and EOC Figure XIV-5-26a 09/19/00
ACTUAL LEVEL ********* ft 1-R SENSED LEVEL. ***** ft hll 9::~0 LEVEL ****** ft Core 8oN>n Cone
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l 0.8 TIME (SEC>
1.2 t.6*10 1 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Transient Response ofLOAP at MELLL and EOC Figure XIV-5-26b 09/19/00
"'" ----------,---~-,----,----,
TIME,HR
"' 1 I
~
35 -'/.,-----+----t------t-----i---~-4
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j
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v+---__ _,_,,__, ___ __,_ __ ~-c+-~~-.:--~--c-<
Nebraska Public Power District COOPER NUCLEAR ST A TION UPDATED SAFETY ANALYSIS REPORT (USAR)
Station Blackout Figure XIV-5-27 02/12/15
-0 (I) 400 ~~--~~----8--'----N~e-u~.-ro-P-. =F~iu_x_{~i-g~t-a-x1-*s-)~ 800
-?l-Ave Surface Heat Flux 350 300 250 a
10
--;1t;;- Core Inlet Flow Core Inlet Subcooling 20 30 Time (see) 40 700 600 50 2/lO ~~-------------~----~ 80 -' Level {rign axis) 175 150 125
~
Vessei Steam flow
-a-T urblne Steam Flow Feedwater Flow 70 60~
t:'.
i:
(J)...
0 50 ~
II')
0.
(I) r,/;J 1DO 1/4-~>d!H~~--ft-i~
40 ~
0 75 50 25 0
10 20 30 40 Time (sec}
50
-§
'ii>
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~
100 -~------8--'--~-o-m-e~P~e-s-s~R~1s_e_r=, -ig=h_l_a_x-is=)-** 400
~
Safety Valve Flow
-Jr-Relief Valve Flow t
360 90 BG 70 6D 40 30 20 (j
10 Bypass Valve Flow
~
HPC! Flow (% F\\N) 20 3G Time (sec) 40 50 320 280
.e, 24G
<1>
IJ) o!:
<l.>...
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160 iv 12G 4C a
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D 4_0 -------------~-~----~ 4.D -' Void Re-adivrry 2.5 e 2.0
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0
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10 20 30 Time (sec}
40 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip BOC to MOCl: MELLLA-HBB Figure XIV-5-28a 01/21/19 50
300 250 150 5D 1 *
B B *
[!
0 10 200 175 150 125
-0.,
1if 100 0::
':!?.
0 T
75 t 50 25
{)
0 10
-B-Net ron Flux (. ignt axis)
-;W;- Ave Surface Heat Flux
-;k-Core Inlet Flow
--+-
Core lnlet Subcooiing 20 30 Time (sec} 4(l Leve! {righ axis)
~
Vessel Steam Flow
51)
-r-Turbine Steam Flow
--+- Feedwater Flow
\\
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10
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Safety Valve Flow
-:k-Relief Valve Flow
-+-
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-e-HPC! Flow (% F\\f\\!)
20 30 Time (sec}
40 36G 320 230 50
,----.----.-~--~~~~~~--~ 4.D
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~
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10 20 30 Time (sec}
40 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip 50 BOC to MOCl: MELLLA-HBB - 1 TBVOOS Figure XIV-5-28b 01/21/19
400 800 100 400
-e-Neu*ron Flux ( ig 1t axis)
--El-'
ome ess Rise ( igl1t axis)
-¥.:- Ave Surface Heat Flux
~
Safety Valve Flow
-;II;:-* Core lniet Flow 90
---ti;;-
Relief Valve Flow 360 350
-$- Core Inlet Subcooling 700 Bypass Valve Flow
-e-HPCl Flow (% FW) 80 320 300 600 70 280
'o' u;
250 soo B
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()
0 10 20 30 40 50 0
10 2G 3a 40 50 Time (sec)
Time (sec) 200 8G 4.0 4.0
-e-Leve! (right axis)
-e-Void React1vrty
~
Vesse! Steam F!O','I'
~
Doppler Reactivrr-1
~
T urbme Steam Flow 3.5
-ii,;- Scram Reactivity 3,5 175
~
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~
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10 20 30 40 50 0
10 20 30 40 50 Time (sec}
Time (sec}
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip MOCl to MOC2 : MELLLA-HBB Figure XIV-5-28c 01/21/19
400 350 300 250
-0
,v
~ 200
?;~
150 WO 50 0
200 175
150 125 "O
2
" *rno n:
75 50 25 0
-~-~~-~~-~~-~~~~~~~~
600
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-,(- Ave Surface Heat Flux
--.w.-
Core Inlet Flow
-+- Core Inlet Subcooling 20 30 Time {secj 40 700 600 50
.-~-~,-~-~,--,-~-r,-,---,--,...,,--,.....,-,,-,--..,,....,~,.......,... OfJ
-e-level (rigri, axis}
0 10
~
Vessel Steam FIG>','
-,&- Turtrine Steam Flow
-+- Feedwater Flo'ov 20 30 40 Time {sec)
!C 50
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"Iii 0::
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10
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-7(--
Safety Valve Flow
-w.':--
Relief Valve Fiow
-+-
Bypass Valve Fiow
-e-HPCI Flow (% FW) 20 30 Time (sec}
40 t
360 t
320 280 50
.-~-~,-~-~,--.----::,-,77::,,-c;"'n"':::-::r:-:;-r-:'lc'--r 4.lJ
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livify
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0
~
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~ I f
3.5 3.0 2.5 2.0 e
/!I) -
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rn E 0
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~
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-1.0 I ' I.., I.,..
I l
I 40 I I Ir~:
10 20 30 Time (sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip MOC2 to EOC : MELLLA-HBB Figure XIV-5-28d 01/21/19 50
,!QO
~
Neu*ron Fiux ( ig taxis) 800
-;t-- Ave Surface Heat Flux
--:&-- Core Inlet Flow 350
~
Core Inlet Subcooling 700 300 250
<I) f;_ 200
,:i O'
150 100 5D D
0 10 20 30 40 50 Time (sec}
2.00 Leve! (righ 80
~
axis; t
mt
~
Vesse! Steam Fiow
~
Turbine Steam Flow
-+- Feedvvater Flow 70 "150 t
i:
<I) 0 125 50 t;;,_
It!
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r.t>
~ 100
~
40 g'.
0
~
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75 30 {3 C:
-a; 50 20J 25 10 0
0 0
10 20 30 40 50 Time (sec}
wo ~~-~~-~*-~~o_m_e_~~e_s_s~R=,i-s~e~(~ig~h~t~a-x--is-)~ 400
~
Safety Valve Flow 90 80 70 40 0
0 10 4.0 I 3.5 3.0 2.5
-.i11r-Relief Valve Flow
~
Bypass Vaive Flow
-e-HPCI Flow (% FW) 20 30 Time (secj
-I-'
--;,(--
~
--.i,-
40 Void Re CINity Doppler Reactivtty Scram Reactivity Total Reacilvity 36()
320 280 240 200 16()
120 8G 4ll (l
50 4.D 3.5 3.0 2.5 1o
£:;
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i
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-1.0
-1.5
-15
-2.0
-2.0 0
10 20 30 40 50 Time (sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip MOCl to EOC: MELLLA-HBB - 1 TBVOOS Figure XIV-5-28e 01/21/19
400 -~--~--~-*~~N~e-u~r-o-n~F=i-ux~, ~( ~ig~,~t-ax--is=)~ BOG
~
Ave Surface Heat Flux 350 300 250 0
10
~ Core Inlet Flow
~
Core Inlet Subcooling 20 30 Time (sec) 40 50 700 500 400 ~
u:
C e 300 ~
z 100 2DD -,--~---,--,---.-----,.,-,-----,....,-,.,....,-,,---------,-,-,-,--..--, 80
-a-Level {right axis)
-x:- VBsse! Steam F!ow
-r-Turbine Steam flow 175
~
Feedwater Flow
?G 150 75 50 25 10 I
0 0
10 20 30 40 50 Time {sec}
100 90 80 70 60 4()
30 G
10
-e-ome P ess Rise ( ight axis) 400
~
Safety Valve Flow
~
Relief Valve Flow Bypass Valve Flow
-e,- HPC! Flow(% FW) 20 30 Time (sec) 40 360 320 280 50 4.0 -~------~----,--~--~---~-~ 4.D
-e-Void Reactivity 3.5 3.0 2.5
-0.5
-1.D
~
Doppler Reactlvrty
---:,1.- Scram Reactivity
~
Total Re-activity 3.5 3.D 2.5 2.0 e I,/) c 1.5 ~
0
0.
1.0 g 0
05 f
g rG
+------------+ 0 0 ~
-0.5
-1.0
-1.5
-2 0 -j--;-+-+---+-+-+-+-+--+"-,'-;-_,_+---!--t--+-r-+-1--+-r-+--+-+ -2. 0 0
20 30 Time(sec) 40 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip BOC to MOCl : ICF-HBB Figure XIV-5-28f 01/21/19 50
400 350 300 250
-0 4>
~ 200
~
150 100 50 0
ZDD 175 15D 125 "O
4>
ri, 100
~
0 75 50 25 0
--~-~--~-~~-~~-~~~~~
800
-B-Neutron Flux (;ig taxis)
I 0
10
---7t-Ave Surface Heat Flux
--r-- Core Inlet Flow
~
Core Inlet Subcooiing 20 30 Time{sec) 50 700 60()
'o' 50D 2 ro 0::
~
e..,,
40D i:i u::
2DD
,--~---r-,----,--,-~-rc-r----,...,,.---c-c-,,--,---,,----,-,......,..- 30
-a-Level (right axis) 0 10
~
Vessel Steam Flrn.v
-r-Turbine Steam flow'
~
Feedvw"aler flow 20 30 4()
Time (sec}
50 70 20..J 1()
1{H}
90 30 70 60 40 30 20 10
-B-ome P,ess Rise ( 'ght axis) 400
~
Safety Va!ve Flow
--:,iir-Relief Valve Flow
~
Bypass Valve Flow
-e-HPCI Flow (% fl/V) 20 30 Time {:sec) 40 51]
360 320 280 "iii E::
240 o 1h a:
,t}
200 5 lb E
a..
160 Q.)
120 80 E
0 0
4, D -.--~--,--.-----,-----.--,-,,-,,-,=-.,,.-c,,-,,-.-,- 4,0
-B-Void Reactivity 3,5 3,0 2.5
~ 2.0
-0.5
-10
-1.5 0
10
~
Doppler Reactivity
-,- Scra'TI Reactivity
--;i,- Total Reactivity 20 30 4G Time {sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip BOC to MOCl : ICF-HBB - 1 TBVOOS Figure XIV-5-28g 01/21/198 3.5 2,5 2.U e i:
1.5 ""
C 0 0.
10 E 0
0 05 =s:
0 ro 00 &;
-0,5
-to
-t5
-20 50
400 ~-~-~----a-~r~'le-u~ro~n~Fl~u-x~(~, i-g~h.~, -ax_i_s~)~ 800 350 100 250
-0 ii;; 200 a:::
~
150 100 50 a t 0
200 175 f 150 125 "O
2 ltJ 100 cc.
~
75 50 25 0
0 10 10
-x-Ave Surface Heat Flux
--.:1t-Core Inlet Flow
~
Core Inlet Subcooling 20 30 40 5G Time{sec}
-El-
~
Level (right axis)
'iesse~ Steam fiov1 T urbme Steam flaw feed'.vater flov1'
,< ;,( H H\\
I I 20 30 40 Time (sec) 50 500 80 70
<ii 20 --l "O
Q) -
Ill a'.'.
~
0
ioo ~~--~~--*-* ~D~o_m_e~P~?e_s_s~R~1s_e_(~f~ig=11~t-a-x~is~)-. 400
~
Safety Valve Flow 90 80 70 60 50 40 20 4.0 3.5 3.D 2.5 0
10
--;;!i;-
Relief Valve Flow Bypass Valve Flow
-e-HPCl flow (% FW) 20 30 Time (sec) 40 50 30 320 280
q;
,8; 240 "' "'
ii::
200 5 r.n r.n f a..
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0 0
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~~-~~-~--1
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-' Level (righ axis)
~
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-:ii,;- Turbine Steam F!o1t1
-@>-- Feoowater FlC¥,
\\
800 700 600 40D ~
u::
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t Jc.,...
0 5G ~
w a.

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0
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<l) 30 £ C -
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--!ii:.-
Relief Valve Flow Bypass Va!ve Flow
-e-HPC! Flow {% FW)
I 36{)
320 280
,w 8G
\\t 4G
_,........,~iiil=lii:~~........,__,J,,...+,....:.....,~,_...-.ll G 0
10 20 30 Time (sec) 40 50 4.0 ~-~~---~-~~-~~~~.,.......,--~ 4.D ' Void Reactivity*
-:X:- Doppler Readivit; 3..5
--k-Scram Reactivity 3.5
--,.... Total Reactivity 3.0 3.0 2.5 2.5
-0.5
-0.5 j
JI::
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-2. (i 0
10 20 30 40 50 Time (sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip MOC2 to EOC: ICF-HBB Figure XIV-5-28i 01/21/19
400
'-a-Neuron Flux ( ig taxis)
BOO 100
-Er ome P ess Rise ( ight axis) 400
-:3/4;-- Ave Surface Heat Flux
~
Safety Valve Fiow
---:ii,-- Core inlet Flow 90 Relief Valve Flow 360 350
-@- Core Inlet Subcooling 700 Bypass Valve Flow
-e-HPCJ Flow (% FW) 80 320 300 600 70 280
'rs 1ii 25D 500 2 E::
ro 60 24G a,
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0..
40 160 0 150 300 ~
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0 0
30
\\
120 mo 20ft (7
2D 8G 50 10 40 G
0 D
0 0
10 20 30 40 50
(]
10 20 30 4{)
50 Time (sec)
Time (sec) 2()0 8G 4.D 4.IJ
'-a-Levei (rlghI axis) -' Void Reactivity
-Pt-Vessel Steam Fl01P1
-Pt-Doppler Reactivity
---:j);- Turbine Sleam Flow 3.5
.......- Scram Reactivity 3.5 175 t
-+- Feec!vvater fkYN 70
-.- Tota! Reactivity I
3.[i 3.D 150 60-2.5 2.5 t:'.
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.... D.5 05 ~
\\
(I) 75 30 'E
(.l 0
i::
<1iS
,a 1~
G>
D.O 00 ~
~
0::
'1}
50 20...1
-0.5
-0.5 I
-1.0
-1.0 25
"~1 10
-l.5
-1.5 0
\\
0
-2.0
-2.0 0
10 20 30 40 50 0
10 20 30 40 50 Time (sec)
Time (sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/LS Turbine Trip MOCl to EOC: ICF-HBB - 1 TBVOOS Figure XIV-5-28j 01/21/19
400 800 100 400
'-El-Neuron Flux { ignt axis)
'-El-ome ess Rise ( ight axis)
~
Ave Surface Heat Flux
~
Safety Valve Flow
~
Core Inlet Flow 90
-Ji,:-* Relief Valve Flow 360 350
--Ir-Core Inlet Subcooling 700
~
Bypass Valve Flow
-e-HPCI Flow (% FV\\f) 80 320 300 600 70 280
'o
'iii 250 500 ~
..3:
&O 240
<1>
a!'.
-0
~
-0 0::
<l>
£_.
l'1)
(l) ci> 200 400 !:;
1is 50 200 5
£l:'.
Li:
tl!'.
~
~
0 C
0 l'1)...
0
a.
41)
'!60
150 300 al E
z 0
0 J'()
120 100 200
\\
~I 2D 30 I
\\
50 101)
\\
1()
rt
\\
40 0
D D
0 0
iO 20 30 40 50 0
10 20 30 40 50 Tim.e (sec)
Time {sec}
WO 80 4.0 4.0
'-El-Level {right axis)
'-El-Void Reactwity
~
Vessel Steam Flow
~
Doppler Reactivity
--;ii;- Turbine Steam Fla#
3.5
-r-Scram Reactivity 3.5 175
--.- Feedwater F!ow 7()
--.- Total eactivrry 3.0 3.0 15[)
60-2.5 2.5 t:
ii:
~ 2.0 2.n e 0
125 50 ~
tJ>
c c
1'11
<I) 1.5 1 5 l'1)
0.
C C
-0
<I) 0 0
<I)
Q.
a.
't;J -WO 40 ~
E 1.0 1.D E oc 0
0 0
~
.0 0
0 O*
<II
"' - 0.5 05 ~
<I)
s:
75 30 o 0
C C
l'IJ ltS
()) 0.0 00 ~
Q) tl!'.
l'1) 50 20..J
-0.5
-0.5
-1 0
-1.0 25 10
-1.5
-1.5 0
0
-2.0
-2.0 0
10 20 30 40 50
()
10 20 30 40 50 Time (sec)
Time (sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip EOC : MELLLA-UB Figure XIV-5-28k 01/21/19
400
-s-Neu ron Flux ( ignt axis) 800 100
-}r Ave Surface Heat Flux

,ii;- Core Inlet F!ow 90 350

--ii"- Core Inlet Subcooling 700 80 300 600 70
'o' 250
. 500 2
'1l 0::
60
-0
/\\
e (I)
~ 200
~~
400 ;
R u:
C
~"
0......
150 30D ~
~
z 100 200 "O
(I) ro 50 0::
~
0 40 30 2D 50 100 10 0
()
)
0 10 20 3()
41]
50 Time (sec}
200
-a- 'Leve! (righ't axis) 80 4.0
~
Vessel Steam Flow

,ii;- Turbine Steam Flow 3.5 175

-@-- Feedwater Fimv 7r.i 3.D 150 60-2.5 t:'.
i:
00...
0 E 2.0 125 50 ~
'-rn Cl.
"O (l)
/j)
I/) c
"' 1.5 C:
0 Cl.
40 g!
'1l *wo er; 0
ae
.a rn 00
(!)
E 1.0 0
0 z.
D.5 75 3iJO C

<.)
rn Q)
Cl 0.0 0::
50 20.J
-JJ.5
~t 10
-Ul
-1.5 0
0
-2.0 0
10 20 30 40 50 Time (sec) 0 T
1 I
0 400
-s-ome.ess Rise ('ight axis)
~
Safety Valve Fiow
-I,:-
Relief Valve Flow 360
--ii"-
Bypass Valve Flow
-e-HPCl Fiow (% FW) 320 280 E::
240 (I) t '"";
I t "" j 120 81.l
=IJ\\
~
r 40 0
10 20 30 40 50 Time {sec) 4.0 ' Void Re ct1vi,y
~
Doppler Reactivity
-r-Scram Reactivity
--ii"-
tivity t
r I
I I
I I
I I
I t I ' I I
I I
I I I
I I
I 10 20 30 40 Time (sec)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip EOC : MELLLA-UB - 1 TBVOOS Figure XIV-5-281 01/21/19 3.5 3,0 2.5 2.0 e:
/j)
C 1.5 (l)
C 0 a.
1.() E 0
0 05 i
g
'1l OD~
~0.5
-tO
-1.5 T
t -2.0 50
400 -,--~~-r-----~--,----.,-.,---..------,=,,--,-r,-c:,--.,--,----,.,--,- 800
'- Neuron Flux ( ig taxis) 350 300 250 150 100 50 10
~
Ave Surface Heat Flux
-,tr-Core ln!e! Flow Core Inlet Subcooling 20 30 Time {sec}
40 50 700 600
~
500; DC
':12.
i?..,,
400 ~
u:
C: i
wo ~
z 200
mo 0
2DO -,-~--~~---..,......~-~~~~---cT.-~~- 80
-e-Level (righ' axis) 175 150 0
10
~
Vessel Steam flow
-k-- Turbine Steam Flow
_.,_ Feemvater Fiovv 20 30 40 Time (sec}
70 10 50
co -,---~-.----=*--..,,""'o-m-e""'"'eC"'.s,.,s-;=sR""1s=----e--,-,(,-ig"'ll"t_a_x7is"')-,- 400
---7t-Safety Valve Flow 90 70 60 40 30 1()
0 0
10
~
Relief Valve Flow
~
Bypass Valve Flow
-il-HPC! Fiow {% FW)
W 30 Time (sec) 40 360 t
320 230 50 4 0 -.-~-~--,----~--,----~--.-,r,-,,--,,-:a;:,--,--,~,-,----,- 4.D
-e-Vo~d R c1ctrvfty
-;>(- Doppler Reactivity 3.5
-,&- Scram Reac!ivity 3.5
---+- Tota! Reactivity 3.G 3.Li 2.5 2.5
-0.5
-1.0
-1.5
-1.5
-2. 0 -+-+-t-->--+-+-t-+--<-f'-+-t-+---i--t-+-t-+-t--+-+-t-+-+-+ -2. 0 10 20 30 Time {sec}
40 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/LS Turbine Trip EOC: ICF-UB Figure XIV-5-28m 01/21/19 50
400
.:....e-Ne rnn Flux ( ig taxis)
BOG
---x-Ave Surface Heat Flux
-:k-Core lnlet Flow 350
~
Core Inlet Subcoo!ing
?DD 300 60()
-0 250 500 2 l/l$
Cl:'.
'O
~
e.,.,
li_ 200 400 :::;
~
ii::
C
~
0 150 31;0 ~
z
~
100
\\)
~~
20D
\\
5U I -~:w.~
ifr;)
\\
~~
\\
JJ l) 0 1D 2C 30 40 50 Time (secj 201i
-s-- Leve! {right axis) 8rJ
~
Vessel Steam Flow
-:r.- Turbine Stearn Flow 175
~
Feedv,ater FkYN 70 15[<
6C-t:
i:
0 125 50 1;;i
'15
0.

I
-:;; 100 40 ~
!l'.:
\\
0
~
I
.a 0
I
,,s 1\\
~
Iii)
'1>
75 30-£
~\\
C "iii,.
50 2(i...I
. r I '
25
\\
1G
\\
0
--t-<--+-
0 0
iO 20 30 40 50 Time (sec) 100 9;)
30 70 60
-0
'ii ri:: 50
~
0
~,.,
..,;:.,i 3D 20
11)
,tO'
-, C,.,
3JJ 2.5
~ 2.D
"' i:
"' i.5 i:::
0
0.
E 1.D 0
0 c
s: D.5 t.:;
\\\\11
<I) 0.0 ft:
-D.5
-'LO
-1 5
-2.0 0
10 ome,ess Rise ( ight axis) 400 Safety Valve Flow Relief Valve Flow Bypass Valve Flow HPC! Flow (% f\\.N) 20 30 Time {sec}
40 50 360 320 28G 80 41}
G
...-~-.-~,-,.-~-,-,~~~,.....,.-,,-,,...,-,="-.-~,,-,,--. 4 i)
-El-' Void RP ctwity t
-.x-Doppler Reactivity
--..tr-Scram Reactivity
-.- T otai Reactivity 3.5 t
3.0 t
25 1
ll.5 t-1.0
-1.5
+->--+-+-_,_l-+-+---<-+-1"-+--+-+-t-+-+-+-t--+--l---+-i--+---t-+ -2. 0 0
10 20 30 Time {sec) 40 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Inadvertent HPCI/L8 Turbine Trip EOC: ICF-UB - 1 TBVOOS Figure XIV-5-28n 01/21/19 50
1200
~ 1000 C.
w a:
w 800 a:
ri.
w w >
600 50 40
-~
e UJ a:::
V)
V)
UJ 30 a:::
0... _,
uJ 3:::
a:::
0 20 10 0
0 2
3 Ae = 0.0147 H = 6 FT TEST OATA CALCULATED x--
X 0
o Tm= 184° F, CARRYOVER ::::95%
Tm= 70° F, CARRYOVER::::-77%
X To;= 64° F,CARRYOVER:::;30%
0 To;= lOJO F, CARRYOVER:::: 8%
CALCULATED HOMOGENEOUS CARRYOVER
- - - CALCULATED ZERO CARRYOVER
--CALCULATED ZERO CARRYOVER AND CONDENSATION 4
5 6
7 8
TIME (SECONDS)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Humboldt Primary Containment Pressure Response FigureXIV-6-1 03/27/00
1200 As" 0.0573 sq. ft.
1200 To= 1500f 1000 1000 U)
~
w a:
~ 800 v.,
800 w
a:
Q..
w v.,
v.,
~ 600 600 0 TEST DATA CALCULATED HDMOGENEDUSCARRYOVER 60 60 50 50 0-ZERO CARRYOVER
~ 40 40 v.,
~
w a:
v.,.,,
w 30 30 a:
Q..
w s:
Ag=AREA OF THE BREAK a:
20 20 0
To=ORYWQL TEMPERATURE 10 10 o.__ ______
__.1...-______ -'--------"-------...A 0 0
2.0 4.0 6.0 8.0 TIME (SECONDS)
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Bodega Bay Primary Containment Pressure Response Figure XIV-6-2 03/27/00
1200
~
.e w a: 1000 Cl)
Cl) w a:
0..
w Cl) 800 Cl) w >
600 50 40
.e w
a:
Cl)
Cl) 30 w
a:
0..
w s:
20 a:
0 10 0
0 2
4 6
8 TIME (SECONDS)
As= 0.021s n2 To= 65 Of 0 TESTOATA CALCULATED HOMOGENEOUS CARRYOVER ZERO CARRYOVER As= AREA OF THE BREAK TD= DRYWELL TEMPERATURE 10 12 14 16 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Bodega Bay Primary Containment Pressure Response Figure XIV-6-3 03/27 /00
~o r--r----,r---.---.---,---r---.--....--....--,---,---r---r----,---r--,
e e
140 120
~ 100 0::
~
w 0::
0..
j 80 w
0::
0
~
E x 60
E 20 t AT 184 psia AH-33
\\
\\
t,.
\\
\\t-*
HUMBOLDT BAY TEST DATA BODEGA BAY TEST !WITH DEFLECTORI DATA FLAG I / 1 IN01CATES CALCULATED POINTS
,_,,k~,':;:~!:'----------t B-26 B-17.30 H-22

----- -,fl,.

0,.__..__..__..__.,..,__.,..,__.,..,__.,..,__..__..__...__.,..,__...__...__...__...___,
10 20 30 40
. 50 60 70 80 90 100 llu 17, 1 IO
!AO 150 160
.VENT ARF'- BRf.'-1< ARf'- flATIC A_. ~
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Comparison of Calculated and Measured Peak Drywell Pressure for Bodega Bay and Humboldt Tests Figure XIV-6-4 03/27/00
CClOPEn cnm>ER CClNT
RESPONSE
CllNl RE51'llN5t USRR CASE A
60.
1 ORYHELL PRESStjHE i
WETWELL PRESStnE USAR CASE A 300.
, ORYHELL TEMPEi HllJIIE 1 SUP. POOL TEM.RATI IRE l!Q, I 1
l.'.)
If
\\ I I
-*(
(.()
Q_
I'-
I
20.
w a:
J CJ)
CJ)
LJJ a:
CL
o. Lw....1-1
0.
1.25 2.5 LCIG TIME - SEC 1216') 1101,11 Cue A NOTE:
I 200.
LI..
(..'.)
I I w 0
I I
w a:
> I 00.
I-a:
cc w
CL
~
w I-I l
I
'>I' 3.75
5.
O F 1 1 1 1 I 1 1 1 1 I
b.
1.2s 2.5 3.75
5.
U*G TIME 121Hl l'IOU Operation of both RHRS coollng loopa-4 RHRS pumps, 4 RHR Service Wuer Booster pumps, 3 Service Water pwnp1, and 2 RHRS heat exc:hangen - with containment spray *.
The short term response curves (i.e., prior to 600 seconds) have been shown on this figure for information_ only. Refer to Section 6.3. 7.1 and Figures XIV-6-1 s" and XIV-6-19 for the short term analysis.
- SEC This case has not been reanalyzed.
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Primary Containment Pressure and Temperature Response Case A Figure XIV-6-5 03/14/02
6U.r*-------,
40.
(.'.)
(Ji Q_
I
20.
w a:
en
([)
I.I.I a:
Q_
w am, 111'8l nto.11 1.25 2.5 LOG TI ME - SEC CaseB NOTE:
COOPER COOPEf{
CONT
RESPONSE
USAR CASE B CONT RE°SPONSE USAR CASE B 1 0RYHELL PRESStjRE 1 WETHELL PRESSl flE 300.
, 0RYHELL TEMPE r11ns 1 SUP. POOi. TEM 'RRTURI:
/"\\ ___ _
200; l.J..
<.!)
w 0
I llJ a:
> 100.
I-a:
a:
LI.I (L
E LLJ I-3.75
5.
0. o.
1.25 2.5 3.75
s.
3"11 CIOTII LOG TIME
- SEC lllCU 1110.*
Opentlon of one RHRS coollDg loopwlth2 RHRS pumps, 2 RHR SeIVlce Water Booster pumps, 2 SeIVlce Water pumps, and I RHRS heat exchangen - with cont.1:lnment spray.
Tlie short term response curves (i.e., prior to 600 seconds) have been shown on this figure for information only. Refer to Section 6.3. 7.1 and Figures XIV-6-18 and XIV-6* l 9 for the short term analysis.
This case has not been reanalyzed.
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Primary Containment Pressure and Temperature Response Case B Figure XIV-6-6 03/14/02
60.,
I I
40.
l.'.)
II v I I
U1 Q_
20.
l.1j a:
)
en
<n w
a:
0..
O.LLI..
o.
t. 25
2,5 LOG TIME - SEC 216t) l'1l,lll CaseC NOTE; COOPER CCICIPER CONT
RESPONSE
CONT
RESPONSE
USAA CASE C USRR CASE C I ' ORYWELL PRESSl1AE 2 HETHELL PRESS! RE 300.,
~\\ I I
I* OA'l'HEU. TEHPE~llJIIE 2 SUP. POOL TEHI 'RRTllRE.
1*-
200.
u..
<.!)
I I
w Cl w
a:
)
I-a:
a:
w 0..
k w I-I I
I 3.75
5.
O E, 1 1 1 I 1 1 1 1 I
b.
1.2s 2.5 3.75
s.
~ cmt LOG TIME IZIMS nu **
Operatlon of one RHRS cooling loop with 1 RHRS pump, 2 RHR Service WtU/l Booster pumps. 2 ~ervice Water pumps, and 1 RHRS heat exchangers - with containment llpt'ly.
The short tenn response curves (i.e., prior to 600 seconds) have been shown on this figure for infonnation only, Refer to Section 6.3. 7.1 and Figures XIV-6-18 and XIV-6, I 9 for the short tenn analysis, This case has not been reanalyzed.
- SEC Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Primary Containment Pressure and Temperature Response Case C Figure XIV-6-7 03/14/02
GO.
40.
~ -
(f)
CL I
20.
LJ.J a:
)
(f)
(n w
a:
a..
"' 00111 21u,11u.,
1.25 2.5 LCJG TI ME *- SEC CaseD NOTE:
COCJPER COOPER CONT
RESPONSE
CONT
RESPONSE
USAA CASE 0 300.,
USAR CASE 0 I DA'l'I-IELL PREsswir
/\\. I I
I I DRYHELL TEHPEiTUAE a HETHELL PRESSl RE 2 SUP. POOL TEH RAT! IRE 200.
l.1..
c.::,
w 0
I L!J a:
J 100.
I-a:
a:
LJJ o_
E w,-
o.
[__~~
3.75
5.
o.
1.25 2.5 3.75
5.
ll<ft 00111 LCJG TIME *- *SEC 12161) 1112,1 Operation o( one RHRS cooling loop with I RHRS pump, 2 RHR Servli:4 Water Booster pumps, 2 Service Water pumps, and I RHRS heat exchangers - no containment spray.
The short term response curves (i.e., prior to 600 seconds) have been shown on this figure forinformation only. Refer to Section 6.3. 7.1 and Figures XIV-6-18 and XIV-6-19 for the short term analysis.
This case has not been reanalyzed.
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Primary Containment Pressure and Temperature Response CaseD Figure XIV-6-8 03/14/02
COOPER CONT TEMP RESPONSE TO l.OCA FOR CASE.£_ ASE
'100
' /')
300 -u t
I
§
',I 5 200
\\
I-<
0::
~
I:!
~
I lOO I,
2, 3,
AAS 00210 0)121 J91l,&
LOG TIME - SEC COOPER CONT PRESS. RESl'ONSa TO LOCA FOR CASE.E-~ASE ss. ~
~~
1, 1
O\\.I AIRSPACE TEM' I
ll.J PRESSURE 1-1-1 PRESSURE.
s,
~o. r----1-~~=--+----+-----+----
~
I
~
~
2s,L I
~I.
l 10.................._..................... _______ __. _____ __. _____ __. ____
I, 2,
'5.
q,
s.
~
1 00210 191M LOG TIME - SEC COOPER I
SP TEMP CCM' iEHP RESl'CNSE 1-1-1 AIRSPACE TEI'?,
TO LOCA FOR CASE-dASE 300, l-------+-------+-------1-..:.-------+-----
200, LI..
,,<.!) \\~'
~
I
~ lOO, I-<
0::
~
w.J I-0, I,
MS 00220 0'11<1 1912.0 CaseE NOTE:
2,
3.
s.
LOG TIME - SEC Operation of one RHRS cooling loop with 1 RHRS pump, 1 RHR Service Water Booster pmnp, I Service Water Pump, and l RHRS heat exchanger - with containment spray.
The short term response curves (i.e., prior to 600 seconds) have been shown on this figure for information only. Refer to Section 6.3.7.1 and Figures XIV-6-18 and XIV-6-19 for the short term analysis.
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Primary Containment Pressure and Temperature Response CaseE Figure XIV-6-9 03/14/02
'100, 300. 1--J
<.!)
,,~
I
~ 200.
I-&
~
I-I-....
I-
100, I,
~2 l OOJSa 17-'\\7,91 1 8-:
ss,
' !\\*
I I
COOPER t
Cl-l AIRSPACE TEl1' CONT TEHP RESPCNSE TO LOCA FOR CASE.F. ASE-
~
2,
3.
'I'
s.
LOG TIME - SEC
COOPER 1
Cl-l PflESSURE CCNT PflESS RESPOOSE 1.£.1 PRESSURE TO LOCA FOR C~SE.F JASE I
25, t---------------1--------------------
~
~*
10, I.
2, li.
-1.
s.
2 t OOISl 17<1,; I
)
LOG TIME - SEC 300 200 w..
I I
§ ~
', ~ 100 I -
0, I,
R.IS 00l8I 0~21 J7*7.9 I
COOPER SP TEMP CCNT TEHP RESPa-&
lol4 AIRSPACE IDP TO LOCA FOR CASE.F
ASE
~
I I
2.
3.
'I, 5,
LOG TIME - SEC Case F Operation of one RHRS cooling loop with 1 RHRS pump, I RHR Service Water Booster pump, I Service Water pump, and 1 RHRS heat exchanger - with suppression pool cooling.
NOTE:
The short term response curves (i.e., prior to 600 seconds) have been shown on this figure for information only. Refer to Section 6.3.7.1 and Figures XIV-6-18 and XIV-6-19 for the short term analysis.
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Loss of Coolant Accident Primary Containment Pressure and Temperature Response CaseF Figure XIV-6-9A 03/14/02
C, C, -
M C, N
0 -
C, 0
S!
7

.,.,~------...

M_.__ ______
N,.___ ______
___.O
~
0 ci O
ci 0
(Aep Jad %) 3l'v'H )l'v'37.LN3IAINl'v'.LNO::l Nebraska Public Power District 0
(I)
~
I-z LiJ Cl
(..)
(..)
<(
a::
LiJ I-lL
<(
LiJ
~
I-COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Primary Containment Leak Rate Figure XIV-6-10
0 N
~ -
00 ci c.o ci N
0
..c w
en ct w
_J w
a::
LL 0
z 0
j::
ct a::
=>
0
,.._ ____ _.__ ____ _._ ____ __,_ ____ __, ___ __.c__.__ ____.,__ ___
___,J 0 0 r---
0 c.o 0
Lt")
0..,.
N0l.l:)V3l:I %
0 M
0 N
0 0
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Primary Containment Capability Index for Metal Water Reaction Figure XIV-6-11
DD FILE: C0048887 w
z p:)
IX I-D I-
<[
IX w
z w l'J A
z D u
.0
~.----+-----t----.-------...__ __ _,_ __ _,
m I s: g u..----~
0-----
~+----+----+----+-
a...
<J s:
0
_J u..
..J w
V)
V) w >
IX D
1-u
<[
w IX 2
~
1-(,f)
~~
<(~
2_J Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
MAIN STEAM LINE BREAK ACCIDENT BREAK LOCATION FIGURE XIV-6-12 08/01/03

,------"T"------,,------...... -.

03S070 A77.:I S3A1'vA NOll.'v70SI ---
0
<D 0
'1)
Cl) w o.r,
!i: ;::
z g Q
~
IX!
N.
L-____
....... :--____ __,o o.r, 0
0 N
o.r, BREAK FLOW RATE (lb /sec x 103) 0.....
Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Main Steam Line Break Accident Mass of Coolant Through Break (10 Second MSIV Closure)
Figure XIV-6-13a 09/29/98
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0~~1.a Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Steamline (Outside Containment) -
DC Power Source Failure (Nominal)
Break and SRV Flow Rates (5 Second MSIV Closure)
Figure XIV-6-13b 03/27/00
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8 TIME (sec) 10 12 14 16 Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Containment Pressure FigureXIV-6-16 03/27/00
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Containment Pressure Response Figure XIV-6-16a 09/19/00
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Nebraska.Pub)jc_,Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
DBA Containment Pressure Response Mark I Containment Program Figure XIV-6-18 09/19/00
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Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
DBA Containment Temperature Response Mark I Containment Program Figure XIV-6-19 09/19/00
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2, LOGCTIHE/SECl Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)
Original Short-Term Primary Containment Pressure and Temperature Response Following a Loss of Coolant Accident FigureXIV-6-20 03/27/00

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