ML23129A306

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1 to Updated Safety Analysis Report, Section Xiv, Figures XIV-4-1 to XIV-6-20
ML23129A306
Person / Time
Site: Cooper Entergy icon.png
Issue date: 04/20/2023
From:
Nebraska Public Power District (NPPD)
To:
Office of Nuclear Reactor Regulation
Shared Package
ML23129A261 List: ... further results
References
NLS2023022
Download: ML23129A306 (1)
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Text

NUCLEAR SYSTEM PARAMETER EVENT TYPES EVENT APPLJCATJON EVALUATION OF DAMAGE OF RADIOACTIVE VARIATION MATERIAL BARRIERS OPEN/CLOSE ANY VALVE STOP/START FUEL ANY COMPONENT BARRIER ELECTRICAL FAILURE MANIPULATE REACTOR COOLANT ANY CONTROL PRESSURE BOUNDARY DEVICE SINGLE OPERATOR ERROR 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 CADD Fl LE: C0048885 08/01/03

CATEGORY OF RADIOACTIVE ACCIDENT ACCIDENT EVALUATION OF DAMAGE TO RADIOACTIVE RELEASE OF RADIOLOGICAL MATERIAL TYPE APPLICATION MATERlAL BARRIERS RADJOACTIVE EFFECTS RELEASE MATERIAL COMPONENT MECHANICAL FAlLURE FUEL OFFSITE OVER- AND HEATING ONSITE DOSE PIPE BREAK COMPONENT MECHANICAL FAILURE FUEL OFFSITE OVER- AND HEATING ONSITE DOSE PIPE BREAK COMPONENT MECHANICAL FAILURE FUEL OFFSITE OVER- AND HEATING ONSITE DOSE PIPE BREAK COMPONENT MECHANICAL FAILURE SEC-FUEL ONDARY OFFSITE OVER-HEATING i2i~= 1---------t-"'1 AND ONSJTE MENT DOSE PIPE BREAK COMPONENT MECHANICAL FAILURE FUEL OFFSJTE OVER- AND HEATING 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 DD FILE: C0048886 08/01/03

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@ ~ 0 (031~8 ~O 1N3J ~3dl Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT <USARl Generator Trip (Load Rejection}

with Bypass Figure XIV-5-1 07/22/96

Plant Res onse to Load Re"ection w/o B ass MOC2 to EOC MELLLA UB

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

Generator Trip (Load Rejection) without Bypass Most Limiting during Cycle 33 Figure XIV-5-2 02/17 /23

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Turbine Trip with Bypass Figure XIV-5-3 07/22/96

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PIO: 4206A Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

Turbine Trip without Bypass Most Limiting during Cycle 33 Figure XIV-5-4 02/17/23

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. COOP ER NUCL EAR STATION UPDA TED SAFETY ANALY SIS REPORT (USARl Clos ure of One Main Steam Isola tion Valv e {MSIV)

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Figure XIV-5-6 07/22/96

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£031~8 dO 1N3J 83dl 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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.cl Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR' Inadvertent Opening of a Safety/Relie f Valve Figure XIV-5-9 07/22/96

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1031~8 dC lN3J 83dl 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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(031~H dC 1N3J H3dl Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

Loss of Auxiliary Power (Trip without Transfer)

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Loss of Auxiliary Power (Loss of Grid Connection}

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Recirculatio n Flow Controller Failure - Decreasing Flow Figure XIV-5-13 07/22/96

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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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SLO Pump Seizure for hp1 CYCLE 28 POWER: 68.5% RATED FLOW: 57.1% RATED 1oo .------------------.--~-1e_L_1tr_o_n~F-lu-x----~ 125 .------.------.,.----,-----e-------.\--./e-s-s-el"'D,--o-rn-e--,P,--r-es-s-u-re--,- 1075

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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

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Recircu lation Flow Control ler Failure - Increasi ng Flow Figure XIV-5-16 07/22/96

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Startup of Idle Recircula tion Pump Figure XIV-5-17 07/22/96

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PIO: 4206A Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

Feedwater Controller Failure - Maximum demand Most Limiting during Cycle 33 with all TBV in service Figure XIV-5-18 02/17/23

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Figure XIV-5-21 07/22/96

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

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

ATWS - MSIV Closure BOC, 76.8% Flow, 3 SRVOOS, MAX SRV Setpoint Figure XIV-5-21b 02/08/16

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c 0

a

-N(T')'::1" 0

ID-u u w w

(/) en UJ w

c 2:

0~

r 11:J!Sdl 3t:fn5S3t:fd 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

0

2'

-"T-;;::;.....-----.,...,~-----1-------1N CX) CX) u u lLJ UJ U') en l1J z: ~

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I'

. 11:HSdl 31:lf1SS3\:ld 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

I 1£UTRON fLUX .........

  • 1-
  • ~VG SUlf f£AT fL X****
  • 1-
  • It-UT 9AlCro_ I ** B/H,,.

CffiE It-LET fL0-1 ******

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JOO.

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f-0::

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f-z w

u 0::

w Q_

o.~ 0.8 1.2 1, 6*10' O<J 102299 600631 1022., TIME <SEC) 1.1s 1 - - - - - - - i - - - - - - - ; - - - - - - - + - - - - - - + - - - - -

1.osi-:----:---1, -------+------+-- ----+-----

(J)

Q_

o.~ 0.8 1.2 1.6*10' TIME <SECl 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

  • ACTUAL l(l'Q ********* fl
  • 1-.R 9'.NSED LEVEL. ***** ft
  • hR 9::IG:D LEVEL. ***** ft Core Boron Cone. ppm/JOO 1s.1--------- t--------+---- -------t------ ---j------

1-w w

LL

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o. o.~ 0.8 1.2 J .6*10' OfJ ..,..,.

102299 1022.1 TIME <SEC)

  • VES9::1. STUM Fl ****** 7.
  • TUlBINE FLOW .******** *1..
  • RV FLOW *.************ *7.

SI' FLOW ***.****..***** x J 50 .1---------!--------+----- ---+-,'-!lf'C":::::-l~/RC :==:lc!:C~Fl':'-OW~.-' 1.~.-'-.'-' ....._,,_;,:-,- - -

JOO.I==~~ ~~.;-------,- -------+----- --+-----

0 w

I-er 50, I-z uJ Ll a:

w Q..

o.

o.

....,,, o.~ o.s 1.2 I .6*10' o<<J 102.."99 102M TIME <SECJ 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

0 &L lO-u u w

Ill w

<n w

z: ~

c,;1- 0

  • 1-
1'  :::1'

_ _. J __ ___J...__.._____L.L.l_,_,_...L..L.L..L..u.o a ci

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031tlij .:ill 1/. M01.:l N CC

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)

1..1.J a:

)

~U"l,;...._-1------+----___;1-+--------. ~

1..1.J a:

a...

1..1.J 2:

0 Cl

-N('f")=r ci lO-u u w w

<n Ill w LU z: :z:

cil- 0

  • I-

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--::t..l..o-_----,N~---.L_-'-

a

  • .J....J....L.L.lc......1...1.....l:,:!CDO 0 0

-* - 0 N CD  ::I' (l:J!Sdl 31:lflSS3tld 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 r

Neutron Flux/% Rated} oOO ~-~----.--.----,---,---,--B---,:c----,'l~e~ss~e~IP~,*...,,-,~Ri...,,e-,(p"'scci,---,

0-- Average Surf ace Heat Flux rtYv Rated) ----<$- Safety vatve Flow (% Rated}

~ Core Inlet Flow(% Rated) -1.:r-- SRVFIOW(%RatedSteamFlow;

-e-

\ **-os*,~**=-

250 Core Inlet Subcooling (Etu/lbm} 450 -

225 400 200 350 175 300

~

.., 150 " 250 i 125 2l a:

<? "'200 100 150 75 - 1UD 50 50 25 0 0 -5
0 +--~--+---+-~+---+----+--t----+----+--t-----1 0 5 10 15 20 25 30 50 0 10 20 30 40 50 60 90 100 Time(sJ Tm>e(s) 200

-e-- le-*.:B fin .ma,,-e Sep Slrtrt;:*

~ vesserSteamRw,2,*i%,Ra.retJ';

is- Tuffifule Sti?:am Ff.il",;,li f% Raw;t."J, 175 -e- Fee@~Rmvn-::.F~;.

150 125 100

~

~ ,:-

-:;; 75 ~ 00 -------------!~-l-----------i a:

g ~_ef

"' a:

50 25

-0.5 0

-25

-50 +---+---+--+----+--lc---+----+---+--+---l -1 0 +----+---+--+---+----+--+---+-----+--+---1 10 15 20 25 30 35 40 45 50 0 5 10 15 20 25 31) 35 40 45 50 Time [s) 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/

11200

}rn~(I 1'llXl 1Qf.C 10CJIJ i) 10(\ -100 tiDO 6 n" A

,BOO iOOO '¥2~[)

~ ~.

~oo A, .Ll h LI

,GOO 1&,(l 20"0 Timw(,~:i 200 . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - , ,X!

. _ SuppresliWln Pool Temp.,ralurB (<l'-'1,1 f')

-,,... cor,Minmf1fkt P1'1i;t,Yf (p;ilg}

rn 17;;,

" 0 rn iii:' 160 Mg 1f! 12~ .e:

I 12 !

{! 100 * '0

! a I

~

I i

I...

7ii A, fl (JI 50.

i 4

25,

?

0 '.l 0 illOO 2001) 3{/JO 4000 GOO!) 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

w L

  • 1-

--.;.-----l-./---+1----1-----i§?

.a .o 0 c:i c:i a D D D D "7 ')J N (D  ::r 03ll:Jt! .:JO% MOl.:l

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w a::.

i 0

w co c::

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w 5

D

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0

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r l

0 I/

(\j J I

-- ...: a::,O

~ D 0 D

"' co lt:JISdl 3t!OSS3t!d 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

l-a:-

.:::u..

rr,--

1 ~

Q. _J -

IJ.JLLJ

<n>>-

LLJI-U.-l-lJ.J >

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t-lf')U u.za:

-LLJLLJ

_JV) er:

w

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...J:XZ

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V,

---+--------------01  :::1' -

0 N

031!:ltl ;JO t. HOl.:1 N ID

_J IJ.J*

a:

..;;<fl,;.--+-----+-----+-+-----4al w

a:

Q.

w

c 0

0

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ID-u u w

V, w

lf')

w LU l:: z::

0

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-=~

- - ~ : r - ~ - - - ~ ~ - - - - - ' _ . _ . .........._._._,_,__,_,_,(Dci

.-4 0 0 N

0 co 11:HSdl 38nSS~d 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

  • 1-l.UTRON rLUX ** , ......
  • 1.
  • AVG SU1f f£AT FL X, ... ,1.

' IN...ET S,llCCQ l ** 8/lbm Cm£ JN._ET FLO-I, .....

  • 1.

1so. r,,------t-------f-----'---j----j-!.......l!.QUUl:Ul~u..4~....,_ ,_;1.L._ _

100. ,-----1------tt--lt+-l-~-jf------t ----_:_

0 u.J f--

a:

~

f--

z w

u a:

w CL o.~ 0.8 1.2 1.6*10' 102299 (0()2£1 0'30,3 TIME <SEC)

  • Cm£ PR£5SU,( *** H.psia

' OCH: PR£ssu,[ *** .... psi<1

  • STEN1...INE PRES ~ D l,psio
  • STEN1...lt£ PRES t h 6.psio
  • t~rRO,CUMPQ< -----io 1.25
  • IO' 1.15 b

llu 1.05 k

(f)

CL I-

. ~

0.95 lo I '

l

o. o.~ 0.8 1.2 J .6*10'

°"

102299 ono.3 (0()2£1 TIME (SEC>

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
  • ppm/ l 00 is. 1 ------ -t---- ---t-- ------ +---- -----l -----
s. 1 -----t- ----7f ------- +;,c= =~~"" =J....- ---

1-w w

LL o.~ 0.8 1.2 1.6*!0' Oil 102299 C002£1 0330.3 TIME CSECl

  • Vf.S!B_ STEN1 fl .****
  • r.
  • TWBINE FLOO.,. * ....... 1.
  • RV fl' .. ******* ...... 1.
  • fN FUJII ......... ...... 1.

1so. 1 -------t------,-----~f--------l--_!_~l,£l,W,~,_.__._4 .u"--'-'-..,11.~._ _

  • H"CI/RCIC Fla.I.......
  • 1.

100.

l I

0 w

I-O'.'. so.

,-z w

w O'.'.

Lu 0..

0.8 1.2 t .6*10 1

Oil 10>299 C002£I 0330.3 TIME (SEC>

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!

~ ~-4

-'/., ---- -+-- --t-- ---- t---- -i---

/'

35 J ' l w

+;-- ---+ ---- +--- - ---*

j i-----,

5

!J i 1.

TIME,HR

~1- j

---4 ---- +-- --~ ---~ ----

¥ zo 1 --c-<

v+---_ __,_,,__,_ _ ___,__ _~-c+ -~~- .:--~

Nebraska Public Power Distric t COOP ER NUCL EAR ST A TION

)

UPDA TED SAFE TY ANAL YSIS REPO RT (USAR Station Blackout Figure XIV-5-27 02/12/15

Plant Res onse to Inadvertent HPCI /LS MOC2 to EOC ICF UB 150 300 70

--&- Neutron Flux --&- Vessel Dome Pressure 1200

---tr- Average Surface Heat Flux ---tr- Safety Valve Flow

-a--- Core Inlet Row -B--- Relief Valve Flow

--fr- Turbine Bypass Steam Flow 60 1150 125 250 50 1100 100 200

=c;-

~ 40 0:: 1050 'ia"

't:J 't:J U)

Cl ~

1v 75 ~

<II a::: 150 ~ 1v e:::,

0::

ii:

  • C

'$ t/1 g

IJJ 30 1000 ii Cl z

<II 50 100 20 950

~

25

\ 50 10 900 0 0 0 850 0 5 10 15 20 25 30 35 40 45 50 0 5 10 15 30 20 25 35 40 45 50 Time (s) Time{s) 150 70 10

-e- \lessel Level

-tr- Vessel Steam Flow

-a- Feed.Water Row

-<:,- Turb,ine Steam Flow

-;r- HPCI Row['3/4, of Po'>f]

125 60 0

100 50

-10 E'

i2 g.

II) 75 40

'C II) ~

ff!

<II

~

0:: 0

.Q

~ -20

~

11:1 30:§.

<II 50 0::

'cD CD

..J

-30 25 20 0

-40 *\~ .* . . . . . . *. . . . . . . . . . . . . .

10 * ~

-25 0 -50 0 5 10 15 20 25 30 35 40 45 50 0 5 10 15 20 25 30 35 40 45 50 Time(s) nme (s}

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

Inadvertent HPCI/L8 Turbine Trip Most Limiting during Cycle 33 with all TBV in service Figure XIV-5-28a 02/17/23

Plant Response to Inadvertent HPCI /LS (MOC2 to EOC ICF with TBVOOS (UB))

-e- Neutron Flux 300 80 -..-----.----,-~--,-~-r----,-- .,--e--=---,-v,-es-s---,el"" "D,..,.o=m-e""'Pr_e....,ss-ur_e_....,.. 1300

-ts- Average Surface Heat Flux -ts- Safety Valve Row

-B- Core Inlet Flow - Relief Valve Flow

---v- Turbine Bypass Steam Flow 70 1250 250 60 1200 200 6' 50 1150

~

Al a: ~

iii

~ ~

GI ,e,;

150 ~ ~ 40 1100 !

ii: a:  :::I C

2

  • VI tlJ

~

i a.

GI z 30 1050 50 100 20 1000 25 50 10 950

_ _ _ _ _ _ _......,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _"--'!_ _ 900 0 0 0 0 5 10 15 20 25 30 35 40 45 50 0 5 10 15 20 25 30 35 40 45 50 Time(s) nme (s) 150 -,---,--r----,---,-~-,--,--, --,----,----,--,--e-=--,\i= 1e-ss...,.e~!L--=-e-ve.,..I- - - , - 60 10

-ts- Vessel Steam Row

-a- Feedwater Ftow

~ Turbine Steam Flow HPCI Row[% of P*Afj 125 50 0

40

-10 E'

ii:

tll 75 30 ci.

'O QI tll ~

l:"

~

GI a: 0 J:I

~ -20

-;<< IV

(.)

IV

§.

(I) 50 20 a:

'ii GI

..J

-30 25 10

-40 0 0

-50 +---+---+----t-+---+---- +----+-l--+----+---+---- +-+---+---+---+-----+-+ ---+----1 0 5 10 15 20 25 30 35 40 45 50 0 5 10 15 20 25 30 35 40 45 50 Time{s) Time(s)

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

Inadvertent HPCI/L8 Turbine Trip Most Limiting during Cycle 33 with one TBVOOS Figure XIV-5-28b 02/17 /23

1200 1000

~C.

w a:

w 800 a: Ae = 0.0147 ri.

H = 6 FT w

"'w

>

  • TEST OATA 600 CALCULATED 50 e

40

  • -~ . x-- X UJ a::: 0 *o V)

Tm= 184° F, CARRYOVER ::::95%

V)

UJ a::: 30 Tm= 70° F, CARRYOVER::::-77%

0...

X To;= 64° F,CARRYOVER:::;30%

uJ 3::: 0 To;= lOJO F, CARRYOVER:::: 8%

a:::

0 20 - CALCULATED HOMOGENEOUS CARRYOVER

- - - CALCULATED ZERO CARRYOVER

- * --CALCULATED ZERO 10 CARRYOVER AND CONDENSATION 0

0 2 3 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

As" 0.0573 sq. ft.

1200 1200 To= 1500f 1000 1000 U)

~

w a:

v., 800

~

w

  • 800 a:

Q..

w v.,

v.,

~ 600 600 0 TEST DATA

- CALCULATED HDMOGENEDUSCARRYOVER 60 60 50

  • 0-
  • 50 ZERO CARRYOVER

~ 40 40 v.,

~

w a:

v.,

w a: 30 30 Q..

w s:

a:

0 20 Ag=AREA OF THE BREAK 20 To=ORYWQL TEMPERATURE 10 10 o.__________.1...-_ _ _ _ _ _

2.0

-'--------"-------...A 0 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

As= 0.021s n 2 1200 To= 65 Of

~

.e w

a: 1000 Cl)

Cl) w a:

0..

w 800 Cl)

Cl) w 0 TESTOATA 600 - CALCULATED HOMOGENEOUS CARRYOVER 50

.e w

a:

Cl)

Cl) w a:

0..

40 30

" ZERO CARRYOVER s:

w a: 20 As= AREA OF THE BREAK 0

TD= DRYWELL TEMPERATURE 10 0

0 2 4 6 8 10 12 14 16 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-6-3 03/27/00

~o t

r--r----,r---.---.---,--- r---.--....--....--,---,--- r---r----,---r--,

AT 184 psia HUMBOLDT BAY TEST DATA AH-33 BODEGA BAY TEST !WITH DEFLECTORI DATA 140

\

t,.

FLAG I / 1IN01CATES CALCULATED POINTS

\

120 e

...e

~ 100 \

\t-*

0::

~

w 0::

0..

j 80 w

0::

0

~

E x 60

,_,,k~,':;:~!  :'----------t

E B-26 B-17.30

--- H-22

----- -,fl,.

20 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 USAR CASE A ORYHELL PRESStjHE 300.

  • _ , ORYHELL TEMPEi HllJIIE
60. 1 1 SUP. POOL TEM .RATI IRE i WETWELL PRESStnE l!Q, I 1 I 200.

LI..

(..'.)

w l.'.)

  • -*( If \ I I I I 0

(.()

Q_

  • I'- I I I w

a:

20.  ::> I 00.

w I-a: a:

cc

J CJ) w CJ) CL LJJ ~

a: w l I I-I CL

o. Lw....1-1 O F b.

1 1 1 1 I 1 1 1 1 I 2.5 3.75 5.

'>I' U*G TIME - SEC

0. 1.25 2.5 3.75 5. 1.2s 1216') 1101,11 LCIG TIME - SEC ""'_..,.,.,

121Hl l'IOU Cue A 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*.

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-6-1 s" and XIV-6-19 for the short term analysis.

Nebraska Public Power District This case has not been reanalyzed.

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

COOPER COOPEf{

CONT RE°SPONSE CONT RESPONSE 6U.r*-------, USAR CASE B USAR CASE B 1 0RYHELL PRESStjRE 300. , 0RYHELL TEMPE r11ns WETHELL PRESSl flE SUP. POOi. TEM 'RRTURI:

/"\ ___ _

1 1

40. 200; l.J..

<.!)

(.'.)

w

...... 0 (Ji Q_

I llJ I

20.

a:

w  :::> 100.

I-a: a:

en a:

([)

LI.I (L

I.I.I  ::E a:

Q_

LLJ ,.,.

I-0.

1.25 2.5 3.75 5. o. 1.25 2.5 3.75 s.

w am, 111'8l nto.11 LOG TI ME - SEC 3"11 CIOTII lllCU 1110.*

LOG TIME - SEC CaseB 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.

NOTE: 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

COOPER CCICIPER CONT RESPONSE CONT RESPONSE I*

USRR CASE C I' 300.,

USAA CASE C 60., I I I I ORYWELL PRESSl1AE 2 HETHELL PRESS! RE ~ \

OA'l'HEU. TEHPE~llJIIE 2 SUP. POOL TEHI 'RRTllRE.

40. 1*- 200.

u..

<.!)

w l.'.) II vI I I I Cl U1 Q_

w a:

20. -  ::)

I-l.1j a: a:

a:

)

en w

<n 0..

w k w

I .

a:

I I-I I 0..

O.LLI.. O b.

E, 1 1 1 I 1 1 1 1 2.5 3.75 s.

o. t. 25
  • 2,5 3.75 5. 1.2s LOG TIME - SEC ~

IZIMS cmt nu **

LOG TIME - SEC 216t) l'1l,lll CaseC 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.

NOTE; 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, I9 for the short tenn 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 C Figure XIV-6-7 03/14/02

COCJPER COOPER CONT RESPONSE CONT RESPONSE USAA CASE 0 USAR CASE 0 GO.

I DA'l'I-IELL PREsswir a HETHELL PRESSl RE 300.,

/\. I I II DRYHELL TEHPEiTUAE 2 SUP. POOL TEH RAT! IRE

40. 200.

l.1..

c.::,

w

~ 0

( f) I CL L!J I a:

20.  ::J 100.

LJ.J I-a: a:

) a:

(f) LJJ (n o_

w

E a: w a..

1.25 2.5 o.o. [__~~

3.75 5. 1.25 2.5 3.75 5.

"' 00111 21u,11u., LCJG TI ME *- SEC ll<ft 00111 12161) 1112,1 LCJG TIME *- *SEC CaseD 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.

NOTE: 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 SP TEMP COOPER 1 O\.I AIRSPACE TEM' I

  • 1-1-1 AIRSPACE TEI'?,

CONT TEMP RESPONSE CCM' iEHP RESl'CNSE TO l.OCA FOR CASE.£_ ASE TO LOCA FOR CASE-dASE

'100 300, l------- +------ -+----- --1-..:. -------+ -----

/')

-u

\

300 200, t I

\~'

'*' LI..

§ ,,<.!)

~

5

',I I I 200 ~ lOO, I-

< I-0:: <

0::

~ ~w.J I:! I-

~

lOO I '

3, s, I, 2, 1, 0, s.

AAS 00210 LOG TIME - SEC I, 2, 3.

0)121 J91l,& MS 00220 0'11<1 1912.0 LOG TIME - SEC COOPER ll.J PRESSURE 1-1-1 PRESSURE.

CONT PRESS .RESl'ONSa TO LOCA FOR CASE.E-~ASE CaseE Operation of one RHRS cooling loop with 1 RHRS pump, 1 RHR Service Water Booster ss. ~ 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

~~ NOTE:

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.

~o. r----1 -~~= --+-- --+-- ---+- ---

Nebraska Public Power District

~

I 2s,L I ~I. l COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

~

Loss of Coolant Accident

~ Primary Containment Pressure 10 .................._....................._________._ _ _ _ ___._ _ _ _ ___._ _ _ __ and Temperature Response I, 2, '5. q, s. CaseE

~ 100210 191M LOG TIME - SEC Figure XIV-6-9 03/14/02

COOPER ' SP TEMP

  • lol4 AIRSPACE IDP COOPER t Cl-l AIRSPACE TEl1' CCNT TEHP RESPa-&

TO LOCA FOR CASE.F* ASE CONT TEHP RESPCNSE 300 TO LOCA FOR CASE.F. ASE-

'100,

!\*

.~

200 ......

1--J I ----

300.

w..

' I I I I

' § *,

,,~

<.!)

' ' ~

~ ', ~ 100 I

I

~ 200.

I-

~ ....

I-I-

' 0, I

I- I, 2. 3. 'I, 5, 100, I I, 2, 3. s.

R.IS 0~21 00l8I J7*7.9 LOG TIME - SEC

'I'

~2 l OOJSa 17-'\7,9 1 1 LOG TIME - SEC

  • COOPER 1 Cl-l PflESSURE
    • 1.£.1 PRESSURE CCNT PflESS RESPOOSE TO LOCA FOR C~SE.FJASE Case F Operation of one RHRS cooling loop with 1 RHRS pump, I RHR Service Water Booster ss, 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 8-: COOPER NUCLEAR STATION I 25, t---------------1-------------------- UPDATED SAFETY ANALYSIS REPORT (USAR)

~ Loss of Coolant Accident

~* Primary Containment Pressure and Temperature Response 10, . CaseF I. 2, li. *-1. s.

2 t OOISl 17<1,; I ) LOG TIME - SEC Figure XIV-6-9A 03/14/02

C, C,

0 (I)

~

I-z LiJ Cl M (..)

C, (..)

<(

a::

LiJ I-lL

<(

LiJ

~

I-N 0

- C, 0

S!

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

.....__..,.___ _ _M_.___ _ _ _ _ _N,.____ _ _ _ _ _ _.__ _ _ _ _ ____.O 7~

0 ci O ci 0 (Aep Jad %) 3l'v'H )l'v'37 .LN3IAINl'v'.LNO::l Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

Primary Containment Leak Rate Figure XIV-6-10

0 N

"! ..c

..... w en ct w

_J w

a::

LL 0

~

z 0

j::

ct a::

=>

0 00 ci c.o ci N

0

,..__ _ _ __.___ _ _ __.__ _ _ ___,__ _ _ ___,_ _ ___.c__.___ _ _ _.,___ _ ____,J 0 0

r---

0 c.o 0

Lt") ..,.

0 0 M

0 N

0 0 N0l.l:)V3l:I %

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

Primary Containment Capability Index for Metal Water Reaction Figure XIV-6-11

D I-

<[

IX w

z w

l'J w

z A p:)

z D

IX

, u I-

.0

~ . - - - - + - - - - - t - - - - . - - - - - - - . . . _ __ __,__ __,

mI s:

g 2 u..----~ ~

1-(,f) 0----- ~~

<(~

2_J

~+----+----+-- --+-

a...

<J s:

0 ..J

_J u.. w V)

V) w

> Nebraska Public Power District IX COOPER NUCLEAR STATION D

1- UPDATED SAFETY ANALYSIS REPORT (USAR) u

<[

w MAIN STEAM LINE IX BREAK ACCIDENT BREAK LOCATION FIGURE XIV-6-12 DD FILE: C0048887 08/01/03

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

03S070 A77 .:I S3A1'vA NOll.'v70SI - - -

0

<D 0'1)

Cl) w

!i:

o.r, z

g Q

~IX! .,.

N.

L-_ _ _ _.,___ _ _ _ _...,__ _ _ _ __,_ _ _ _ _.......:--_ _ _ ___,o o.r, 0 0

N o.r, 0

BREAK FLOW RATE (lb /sec x 103)

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

COOPER NS 1 BREAK 1 BREAK 2 t

OBA STMO BRK a BREAK 3 BATTERY FAILURE TOTAL BREAK FLOW I I I

  • 1 1 *5 - * * ~ 1*SRY* ADS xl OI
1. ir------ ---+--- ------1 ------- --'I. **--**-

u LIJ (f)

'co

L

_j O, 5 H-1-S-*, s 1 5

  • t..s...LL1.t-..._s.l-1 ...L~~l..s-s.t.Lli-Sj .1~1-1..!..LLL&**' --'..s-SJ.~-s....s..s.. i4 -1/4..1:..LL1.4 l---\~-1-------- -

uJ I-a:::

I-3: f--

0

_j I-LL. f-I-

00 ~- I ~lt~~ltltltltltftltltHH ltltltlt1tltlt ltltlt It ltlt 1m IMlt It It It It It It ltll It It ltlt ltltlt Ir l!l~ I I I s It e a ltltluu1 II II II II 111 I I o.s 1. 1.5 2. XI O4 LIP 11189&

022EO 0~~1.a TIME (SECOND)

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

50,--- ------- ------- ------- ------- ------- ---

50

'NED0-10;n9 BLOWDOWN FLOW RATE 40 O>

]

w a:

(I)

(I) w a:

Q.

30 I-z w

E z

ci:

I-z 0

u 20

~WETWELL 10 o----------'----------i....--------...&.----"----""'

0 2 4 6 8 10 12 14 16 TIME (sec)

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

Containment Pressure FigureXIV-6-16 03/27/00

~ ~

,  ::, 0 II) II) (")

II) II)

(I) Q) a..

L.

n.

L.

1 Q)

?:

c Q) 0 3:

I l!)

N 0

N 0

Cl) 1/)

Cl)

E r:

l!)

~

I

' 0

~

~

  • 1 l!)

-... "~

~

...... 0

~ ~

0 0 0

~

0

~ ~

0 it it0 0 0 it it 0

0 0 0 it0 it0 it 0 0 0 0 co U) ~

0 M

0 C'li 0

~

0 d

(61sd) eJnSS8Jd Nebraska Public Power District COOPER NUCLEAR STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

Containment Pressure Response Figure XIV-6-16a 09/19/00

0

'It r<>

vN I

'I~ 0 ro I

J I

I j

/

i/

--I I

~v I/

/"

c,,

..,c:,,

"ill C.

I C.

I \

w a::

\ w a:: \

(I) \,  ::>

(/)

(I)

\ Q

~ w a::

a. 8: I

..J ..J

..J w

..J

\

'\

~

~

~ >-

a:: ~

'?..., 0 ) ~

\

_l/

~

- L-- ~

~

r r-....

i'-. ..

..... r-,....

0 r-,...

i...... - 0 0

V PRESSURE (pslg)

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

i c

N

~ ..

IL. IL.

0 0 I I ILi w

0::

<(

0::

I-

<(

0:: 0:::

ILi

0. w
e

...w 0.

e

~

I

..J ..J

..J w

...w 3:

3: *.

Q

~

N

' ")I I

\

j II,_

I ....  ;

~~--

I *.. .... 0 0 0 0

fQ TEMPERATURE (°F)

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

1 DRYWELL PRESSURE 2 WETWELL PRESSURE l DRYWELL PRESSURE 2 WETWELL PRESSURE 60.r ----- -,--- ---.- - 300.i ------ --- ---1-- --l itO. i a:

200.

LI..

(.!)

w

(.!)

( fl a..

--'w 0

a:

w 20.  ::,

a:

I-a:

en a:

en w w a..

a: w

~

a.. I-0;2, o. 2. -;2. 0.----- 2, LOG CTI HE/SEC) LOGCTIHE/SECl Nebraska Public Power District COOPER NUCLEA R STATION UPDATED SAFETY ANALYSIS REPORT (USAR)

Original Short-Term Primary Containment Pressure and Temperature Response Following a Loss of Coolant Accident FigureXI V-6-20 03/27/00

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