ML20024C573

DHR During Very Small Break LOCA for B&W 205-Fuel-Assembly Pwr.

ML20024C573
Person / Time
Site:	Three Mile Island
Issue date:	01/31/1978
From:	Michelson C GENERAL PUBLIC UTILITIES CORP.
To:
References
TASK-02, TASK-06, TASK-07, TASK-11, TASK-2, TASK-6, TASK-7, TASK-GB GPU-0323, GPU-323, NUDOCS 8307120825
Download: ML20024C573 (34)

Text

. __- _ _ _ _ _ _ _ _ _ _ - - - - - -

, , , , . , _g W

y 3 23 (

. ,- ,-~. 6 , .

.-i .

i

, ,. .'5 s '

(

, - e 1

e s  !

, 1 *

• g

, ' e. -

e

)

, t s- #

1 7* f- g 3 \. '

, , . . . - +

~_. ,- , , pq- -

z -

'*+..+G_,,.*. ,

l _'S:y. y . , ' . .  % ':

~

,  %... .. .-t .

• t. .

q 9 - , ' .?', a , -' . ..

. 4 PM g D

- s -~

w. . . . A. ~ , -l*._

te e

/ 8*I e ,*

h e

e b

8307120825 780131 PDR P

ADOCK 05000289 HOL

" ', . :a

'i..

lheSt C0py fi .

. . &. . .s . . ,:

,i si .

. . ..., l . ,., ~ w e p :- c... .r.. _ . . . +s- .s-.

' ' . . ~ ~ -  :

s.t.

. .%' .- ': . . . . nEn-.unT_uram.ca:s a vm s".tt .

.ic-

.. .: 2.

• fy 33rAr tecA rom A Meu 70s.7urt-Assemnty PwR . p,

- c. Michelson p -

. = cory, m . .

.. ..~

.J 3

--y w

.)..

. . q.

This report gives an ac=ount of sore initial considentions of a .-

j . .

]

class of very small break 26's (probably 5 0.05 fta) for a 35W t..

.V ,

205-yuel-Assembly FWR which may have an associated decay heat l'*

• d]

LK, 1

removal problem. The results indicate'that one or more  ! .a .

.] ispediments to decay heat removal appear to exist which need to ,

{

-3 -

?

]i 'be better understood if proper operator response and adequate

mitigation are to be assured. Of particular concern is the . S acceptability of intermittent natural circulation ,following the s postulated LCCA, and system repressurization following the , loss of natural circulation. Also of concern is tige possibility of

. . 9 N* break isolation by eperator action resulting in repressurization h.S ..

']-

and, slug o.r.two-phase flow through a presurizer s'afety. valva.

i

.These uncertainties may reflect on the adequacy of proposed r.j .

emer,gency operating procedures and, operator training for a very

• j small break M cA.

~ ^

.),

-.s........ , -

..: . . n . . . ... . q .: i.* . :....

~ yo .. .s.3 93 u - wo .

. m. .e .! vr xi J.R. Denyo Gya.r.a%.~. W ,%y

. :. ., .+ . . . : . . , . . ... .

.t -

< . ..; .. .. 8 .. . ..

.n .

.i . . . Y 4.

l. - _

.3

j. o 12s . " >""g0-

! [!3j-/Rp_:.J;J.i . .

. l

?'

• .'.Y.,....

beSt

• . C0QY . ,. - . .

a*.

, . . .e-= e _ , , - . , ,

~ .n - .:- - .__

= _

r. .s i .j *

...w.-

.y . . .

. c- .

..s...* n* .. .3:.

- d';.

• s s . .. - -

c ,.. ..

.r.-....c.

s . .. . .

- . ..s 2- - ..- . ,.w : .. . .

. . .. .: corrrENTS

.. . .! r .' .. ...;.::. t -t. . s. g . . -:.

.0,,yf

-r, 3

9

1.0 INTRODUCTION

2.0 1.0cA CRIR.ACTERISTICS .

4 2.1 Mass riow Rate Through Break . ,

2.2 Decay Heat Removed Through Break r.

1.3 Reactor vessel Top Plenum Drain Time d 2.4 Steam Generator Drain Time 2.5 steam Generator Refill Time ',.

i i 30 MODES OF POST-Loch DECAY MEAT RDt0VAI.

• 3 3.1 3.2 Natural Circulation

-./

Transition froo natural circulation to Pool Boiling '

I 3.3 Pool Sciling '

I'i l

2. 4 Transition from Pool Boiling to Natural Circulation 3.5 shutdown _ cooling .

y 1 s.O weRsr case z0cA consIDIRTAr0ws ,

u I i 4.1 Discharge Co.fficient and Break I.oc'ation -

C.

Decay Heat Removal 4.2 j .

s.3 1.evel Turnaround and Energy Equilibrium 3 9 4.4 Usss vendor calculations 4.5 areak Isolation and Pu.p Shutoff Effects 4.6 Pressuriser I.evel Indication

.I .

5.0 con:LusIOus r

. .-.s TABLE .... .

. e- a . ,  : ,, . . .

.l , .

' FIGURES ... . . .

,r< . ,

.e , . . . ,

, ....".o..

+

. .- c .-

..~.,-.:. .

.. . . .4 g; .. . . .. . . .

.- ..x

... . . .r. . : , . :. . , . .  :..

.x.9 . .

. q .. .4 .. . . . .- a

. . v . .. : .. ,:.c . V .. .:..

. .a . - . . . ...; ..

. u . ., ,

l ,, .

h n '. _ ,.. ?.v%: ; 4 . :. : . .

... ..u . .

r.. 4. , . _ _ , _. . ;ey-o

., . g,.c  :

. . ,.,e .:.

. .. ~ .;.s...c .: :-. . .".. <. :, ..s..c.

,.a.... .. . . . . .

.. . z,. .

.2 .... ..

. ....c ..

.n- w.

o

" d'6CM'

. IH3 @ L. W. . -

I, c12s -

, . _ . _ - , , , , . , . , . . _ _ - . . , _ . _ . . . ..,,.m. . , . - _ ~ , , . . - . , _ _ , , , . , . . . , . , ..__ m,. _ . . , ,

E .

J.-

. m . . Lbest copy ..m

. . ,.. . . _ . . . - . .:n: i .._. - .'

,a 3

• . . . / ,

p(  ;... * , N . - - . . ,..n - r

. r, *i.sg

. - ~ .

.v;.-m.v. . .. o p;, ,... .,,, 4 c: . ,. > e .V. ..:. . .

r 1.O Ilr!RCDUCTIOtt '

• k;&.. .

'.. c '

c..

c

.- There appears to be a class of very small break 1.oCA's  ?

(probably 5 0 0.05 fta) for a UW 205-Tuol' Assembly PWR which may- .

D have an'asso'cLated decay heat removal problem. For this 9,'

u

,> 1 '

I discussion, a ,very san 11 break Loch is one for which the steam I

. generator must remove a significant portton of the decay heat j during the initial phase of blowdowns otherwise, reactor coolant l

, . . ) .

] system repressurizatica occurs since the break is too small to

,y% facilitate the transport of all decay heat to the environs. For this class of I.00A's, depressurization rates are relatively slew

.p u

• (when compared to these normally analyzed as small breaks)'and 1

1 thus may seriously limit the makeup available from the high pressure injection ponps. An ongoing qualitative consideration d of this problem now predicts the developmenc of one or more

} .

d impediments to decay heat remeval during a very small break Lc:A. ,

This has 'become a concern that needs to be understood. -

.. ; The physie'ai arrangement of the reactor coolant system for a . . -

n typical 205-ruel-Assembly plant such as the TVA Bellefonte f ,

'i Ibclear Plant is shown in Figure 1. Plant elevations corresponding to various points in the reactor coolant system are indicated. De=ay heat removal considerations during the post- -

1ACA period are based on the usual rect rules such as loss of .

offsite power, minimum core cooling tone train) resp'nse, o and no short-term required operator actions. ,.

.3 . .. . - .."

. ..s.

3 ,

-..t: ,0 j.. .;. y. " ".

v. ' 3 . .

.1 ..

n .. .e.

. ,:~~N ' - . -  :.

s.. .:. .-

... .m. .

y,

? .-

i o12s 3r%o "

! Il3bM?l.J@. ~

. l y:_ .m copy ..

.m f x . L_best . .

a

.f.

'I)

. 2.01oCA Cl!ARACTERISTICS ....

, . c- .

.t. *

.. a ,

hD .-

A few elementary calculations can be helpful in developing a

'h better understanding of t.he various; modes of post-LOCA ~

~

decay beat  :'

a

$ removal and the role which st am generators must play during a .,

' i.

1 i

very small break I,ocA. of particular interest is the mass flow .

rate through a postulated break, its ct.Nbility to remove decay ..

?

i heat, and the makeup rate to the reactor coolant system required R to cor.pensate for any lost mass. Also of interest is the time

. 1. -

,; Id ..

.c required to drain the reactor vessel top nienum and steam g

generator tubes during a transition to pool boiling, and the -

. 1 mininum time required to refill the steam generators if a level i turnaround occurs. i 2

s I .

I For r.ost calculations it is pssumed that a quast-steady-state -

condition exists with the reactor coolant system at 1270 psia i

(

l

- (57a.4*r) and the secondary side of the steam generators

Where relieving steam to atmosphere through a . safety valve.

applicable, it is assumed t! at a sufficient temperature ^

' differential exists across the steam generator tubes to transfer -

,. 4

+

-R.

all decay heat nu -removed by the break. The decay, heat is based i.

.~.Y on the ANS decay heat curvet usin; a 20 percent siargin.

1

@ The fluid upstseam of the break . - .

The'. reactor power is 3672 Mwt.

l t s.-

i;yya The flow areas

's

,,. is N assumed to be esturated water or st;wam.

.a yQ.g. (potential single-ended flow break areas) corresponding to a

' .8 4 number of nominal pipe sizes of interest are listed in Table 1. ,

.I

. . . . . s,',..

. . . ... . n +; a. yw . . . . . . , . . . .

.p 1 .,

5.f ' '* ~ 6.T ~-

• f. . .

~

..=, [ '* ',;)

j

.i

~-

'+.

c 8 Proposed American, 9 I: clear Society Standards

" Decay g

Energy Eclea u Rates rollowing SM tdown of Uranium -

}.* 3

Tueled Therm?.1 Plants" (approved by Subcommittec Alas-5, AliS f.tanstar.is Committee. October 1971).

o ..

4 U'A0'1 (i.!s_W#.JV .

!. 012s

-y ,,e----w w-- w-=, m- ,-ve.i-,ew--e-r- wye--+was--+.yewy=e-y-- -yv------+c.-eeg-p---e---m--u-em-- m-v e r -eem- -+w.,.w-p- -,.pw< w,e--ww-w-_---w-tvy wg -

... copy

• r _ ..

.c.

1 .0est _

f . ,

=

h .

n l

'

• The pipe schedules indicated are typical for the reactor coolant l . . -

' D-c #~. ** ' .

Q pressure boundary. 3,,.

, :. . g; s . . yy. ..

n:s, q .. ? .. .y :.3...V:; ..: *: ;

.g

• It should be recognised that calculations included in this p.j i

P s

sec'.lon are based on general considerat.ionn.cf energy and, mass Fine structure conservation *uncer ideal fixed conditions. -

effects have been ignored to facilitate simple hand calculations.

f ( Bowever, the e suits should still be useful for general guidance f

p if the simplifying assumptions and calculational limitations are

@ J N,1 ,

(

g appre= lated. Detailed transient calculations based on r.

• 9 . .

appropriate system heat transfer and fluid flow models and core f

U~

thermal. hydraulic models are required to put these concepts on a i

j* .

- firm basis.  ; >

4 - 5 6

l 2.1 gges rioJ Fate 'Throuc5 Brea1c l -

k The mass flow rate through a break assuming saturated water er .

L stesse upstreas of the break is shown in Figure 2. The saturated.

I water and stea- curves are based on Moody'st rigure 3 6 for stagnation pressures of 1270 psia and 2500 psia with

. 3 saturated liquid and vapor entrance properties. Moody discharge '

j coef ficients (cp = actual flow / Moody calculated flow) of 0.6 and The 1.0 were sele =ted for calculating the saturated water case.

actual coef ficient might be somewhere between as determined by f such considerations as the break ca.sfiguration and whether it is in a large or small pipe. The discharge coefficient f or steam is assuned to be i c .- .

• • h- -T ' -'n.

? b w. .

.Q .[ -. .'. .

a . s.?..'.$. -- h' 'N -' ..

i i er. J. Moody, "Maxir.um Flow Itate of a single component.25 IIdDHi m IIcD aCI.1023 Two-Phase Mixture," dO!!IDJ1 Of 1!

? 87,840* l' c1 lb2 f. !*IiG2D E2C1C12 Di hC.!!1DiC21 EC3120::I24

+

February, 1965. -

a-..

o12s W '.,% E

!.!! 3 b.E A . M5'

^ --- ,

1 e

-,ww-,, . - - - . ,-,--,--..------..-,.,---,----..,y-w- #-ww,,,.,v-,w,-w-.--, -

---,,,-.--,,....-.,..-,,,,m,.- ~,.w..

(_U t'b L UUpy J- _.; . . . , - .

g. . . .

. .. 4 _ _.

~ -

• . 4 Y y

"' W

• T R~~*:'t.','- ~*

. ai .s.w ' . s n. . ,.-

. .. .: a. .. .  :' ~ ' ~'. V* Y-Y ~ ~~ '.**~ ~5*~ t' G.

y The makeup rate available to the reactor coolant system to ' j",.

g .a

. ,y a.

y.l,'. compensate for mass lost through the postulated break is shown in '

j figure 2 as that available from one high pressure injection (MPI) pu'ap at the dadicated system conditiens. The water available for -

) makeup is assusred .to be at 70* F. .

., ~

v-

.W .

r..

m . ,

7 .

,[~ I'. ,

. 2.2 Deesy gget p e ttroueb a Break .

y.

t; gf . .

The decay heat which is removed by mass flow through a. postulated

.s ,

break.. at. 1270 psia .. (assuming saturated water or steam upstream of

. . . ;g',

, ,3 .

the break) is shown in Figure 3. The heat removed is based on -

the mass flow rate given in Figure 2 and a stagnation enthalpy corresponding to the designated upstream condition, i.e.,

g ,%

• saturated water or stean. A minimum value for Cp (0.5) was J i ',g ,f . selected to yield a conservative. (ninir.ur.) value fo: the heat' removed. The decay heat removed is shown as a perce.st of tota 'I

\* .

~

decay heat generated at the indicated time following the break.

j It should be noted that saturated steam upstream of the ,

(,p q ti * m$' \ postulated break will remove a little larger percentage of the ..

f} ,

N '

decay heat than saturated water. This will not be the case for -

,A '. / '

larger values of CD with saturated water upstrea:n. ' .The larger l .

5$g,*J.

'* 'J mass flow rate for water can z. ore than compensate for ths dif ference in saturation enthalpy at the indicated upstream

- , Y.

' pressure. . . . . ,,

y .

h e time after a break before all decay heat can be .removed i -

t 6,

through the break (with saturated upstream conditions) is -

f

, . .,.] ,

, indicated in Figure 4 This is the cine requLeed for the break

, to be in energy equilibrium with the <*.ecay heat. .

The effect of g.?- ,

S .

' S @.^'

i l i ilsb57. 2@ .

012s

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

.# '~i copy }.. r .

L.oest

, - .. ~ s. - _ _ p .-

.4 E

• 4* y

-- - -- R q.

f...' .;.* . = . ~rc ,e: = w .:

. .y. , ,_. . .a. . . . . .

y

.c~ -

- higher pressure and a range of possible cp values for water is .-Q

%r

~- also shown. - ,

rdi 1 2.3 Eggggg vesset von plenuc.2rainJ ct 4 3.:.

'a i~,'d.

4 The time requited to drain the reactor vessel top plenum down to RF-c.*

the top of the ' bot leg pipes is shown in Figure 5. The drain J

V_

.i '

time is assumed to start after reactor trip.* Fressuriser level [4 .

j r.I r

~ ~

E is assumed to remain at the 1e' vel achieved innediately after ,

i 5 trip.- The pressure is assumed to be a constant 1270 psia. In /

1 hO reality, the pressurizer pressure will be somewhat higher during T

d a portion of' the drain time thereby reducing the drain time 3.-

indicated. Although some additional pressurizer draining siay  :

d, g occur, it is most likely that a refilling will comence shortly ,

y -

,s f ,

after the reactor vessel steam bubble starts to form and become  :

controlling. Therefore, the drain time given in Figure 5 is thought to be a good estimate of the maximus time that naturat

circulation can be sustained for a given break size after reactor

,$ tripe.

- t -

.... ,7 1' . ' '. ;*

2.4 Steam Cenerater Drain Tig . . . . . ,

The time required to drain the steam generator inlet piping, p

y plenum, and tubes doen to the secondary side water level is shown in Figure 6. This drain time is calculated to start when natural .

i.e.', when the

, circulation is assumed to be lost indefinitely, j water level at the top of the stean generator drops to below the. .

inside diameter of the U-bend high peint and restoration of level.

. a.

is not expected. The drain rate is assu-cd to be the w. ass flow a System makeup is rate through the postulated break (TJgure 21 ,

from one high pressure injection (!!PI) pump delivering a constant l

/D-

'.}'. _. _ . .

sy" #

_ . - . . _ _ . _ . . - . . . ___ _ - - , , . .m._ _ . _ _ _ . - . , . - . . - _ , - - . _ _ . . , . . . , . . _ _ . . . - , . _ , , ~ -

m.. .. . . _ . _ !ybest

- c0py ^:- - .:. u.. ...u,

. .. ..r , ..

. w, ,.;

~

- .=

.. . . 6 .

-g g..

, .g. _. .

. ; 9. / ?.e. . . , ..... . , . , . .y . , , , . , , ; . . j-

• N ss flow at the indicated system conditions upstream of the ,

.y .

-) .

break. The secondary side water level is assumed.to be_

k. . i)

A increasing at a rate of 1.75 f t/ min which is attribut;,*d to a 600 -

6:

1 ,

g. ], ~O r . .-

e. ,

gym auxiliary feedwater flow. The initial secondary side water .,y l level is assumed to be 26 ft. above the bottom tube sheet when  ;/.:'

l e steam generator drain is started. .

f' 4 d g .?

3

. A-*'

,.j The effect of, upstreaa pressure and initial secondary side water .-

level on steam generator drain time is also shown. An initial- Y.

.+  :

~9

, (/ secondary side level of 6 ft. is the normal auxiliary feedwater i ,5 *

}

l l control point until an engineered safety features actuation 9: ( t j \ signal (EsFAs) is initiated at 1600 psia. It is likely that the f d

' ' ~

[ steam generator level will be in the range of 6 to 26 ft. at the J

time natural circulation is lost. The effect of an initial level .

y 3 of 48.5 ft (top of s:". shroud) is included. This is an upper L *.

b h limit for practical consideration since the overflow enters the ,

i main steam lines. The effect of 2500 psia reactor coolant pressure (set point of pressurizer safety valves) is included for  ;.

' reference. . , , , .- .

j -l

. ~2.5 Stasm.G;ngratnr RefiMigg n The steam generator refill time is the time required to refill

'the inlet pipe and plenum and steam generator tubes if a level turnaround should occur when the, primary side water level first '

j . reaches the secondary side wat.et level. Typical times are shown

.. in Figure 7 for saturated water at 1270 psia upstream of th's break. The initial steam generator level is that level which

,;_. existed when the steam generator drain r. tarted, i.e., when -

~ ~

.j natural circulation was lost. It is likely that the initial y g.r-g

.. .r g 3 t " 9 ~~N

.[iSh...E.& v.

a i 0126

"' ' U' r b::/'

, - - - - - . - . . . - , , - - , , --,-,-,~,-,,,-..,,---,,-.-,,-,-n, .,-,av-v-, .,,--n,,,-,,..v-,,-.~

~

- r . . . . . . . .

^ 1

_ -L. best copy . . ' . , . ' . . .... .. _ _.

o: . ._.

,t.

j '4 . . . 7 -

+. ... .a..:.....;...,.... * . . . ... ...-... . . .

..)e, .~ .- *

- . , . .. ,. f '.; .

steam generator level' will be in the range of 6 to 26 ft. The y,:;.f.

• ,.:4.Y curves all start at 12 minutes because this is the minimum time *%

. ',c .

} .

j zequired to refill the inlet pipe and plenum and that portion of t

the steam generator tubes which is above the shroud.  :/ a

,x ,.

f.

p .?,. .

,4 .

% 3.0 MODES or POST-1DCA DECAY MEAT REH3 VAL j,i 1,

1

.1 Af ter the loss'of offsite power 'and reactor cool' ant pump [ .' ,,

- g .:2 . .

p

~

coastdowg, t'he two ba' sic stodas of post-LoCA decay heat z'emoval d from the reactor c'ere for very small breaks are nat' ural ih '.

h

. circulation and pool boiling. Decay heat is removed from the g

e reactor coolant system by. single or two phase blowdown through -

J the postulated break and by a release of steam on the secondary I

side of the steam generators. Any de=ay heat lost to the I ,

-$ environs through thermal insulation is small by comparison and is  :

neglected. If the break size is assumed to be sufficiently g

large, two or more of the following phases of operation may be p.

. t; .

experienced. -

c. .

3.1 Natursl_Cireviatio"n - -

. t.

Natural circulation starts with the pressuriser stil'1 controlling

! system pressure at approx 1.mately 2250 p ia. However, without an ,

j effective heat input (other than the pressuriser metal) to .

1 J

compensate for heat removed by flashing the pressuri,ser 11guld. .

t -

1 the pressuri:er level and pressure decrease rapidly as the steam l .,. .

s -

bubble expands to fill the void created by fluid lost through the '

g break exceed'Ing 11guld mak- p capability (1.00A condition).

t . .

j .

rigure 2 indicates a break area execcolng 0.035 f te could create this initial condition at high pressure. ,

r i . .

d' r" . I ; .0" I C 3h.BU[d.T_9 t

! O12G ,

~

• 9

..y_- .,,_ , - . - - . . - - - _ , . , ,..,,.,,,,y.-- . . .1--,,,,,,-.,,.-r------,.--._, -

,,_-,.,-,c- ..__,%,,_.,. --,~,-m.,

I .

,- ~. .

u-..

best C0py Lo -' _. m.:*g,: :._._. -

_,,_3.,- -

g -

~

j  ; - . .. .-

e_ < ce..y.ay . .c.

. . ,.(. r, . . . ,. . . . .

s.3..

. ' ~~ -

.n .

T A significant aysten volumetric contraction and corresponding

y.

.:: -  :-[n. .

i pressuriser bubble expansion occurs when reactor trip is [..

4 initiated at 1945 psia. The bubble superheats to the pressuriser .E

• g-

,j effective metal terperature. The bubble may expand until

. .I-W t; superheated segeam is voiding and condensing in the vertical hot

. .z .

1eg pipe of the reacter coolant system. The engineered safety ,,

.) j 4 r.'

j 1

features actuation signal (EsrAs) initiates at 1600 psia thereby ..

assuring maximum availability of one high pressure injection :t.

vy

?ft*

3 plPI) pump without flow control, and isolation of the letdown . >:

-4 * @, .

- ' system. .

1 A new steam bubble forms at the highest temperature /lewest.

E' l P[ pressure loca' tion in the reactor coolant system when the l

l pressuriter pressure becomes less than the vapor pressure of the a'

liquid phase at the new location. Since there is no significant

i heat sink in the het leg pipes between the reactor vessel and t D stear. generators, the highest terr.perature/ lowest pressure g

location could be in the U-bend pipe at the top of each steam -

' f, generator (Tigure 1).

,. g

~

}- If a steam bubble forms in the U-bend, it interrupts " natural ['

-!kcirculation.

Water in the reactor vessel starts to boil

. c '

' agressively due to loss of circulating flow. The reactor core

,1 c (

a

% y,qA exit temperature and corresponding vapor pressure increase until

! a

~

\ the reactor vessel top plenum be'comes the controlling high temperature location for steam kn.bble formation. This occurs ;...

when the top plenum temperature is about 1 to 2* T above tt.e U . ,.'. .. .

bend te..p:rature. -

. a ... -

. ... .- i . .. ,-

j-l _

" ~

}3].p. ,6,6. b -

j O12G N- -

b v- -..,--n -,_-.--,.m...,--, -

, _ _ , , , , --,,.-,.--.m.--,-,-,.----y r,, . - , , , - ,.--,.-,.p.,%.,,,.,,,_,-- ,,_ . y- - _ , . _ . - , , - - , - - - - -

1

, 1 1

l u.

^

m best c0py 1.... - . : , a .,. ... : .

.._..r,,..

r .

& . . 9 '-

..'y.... , . . , ,

, . . . ~. : .. .. . . . n . 7 ,:- . .: . . .. . g.'

'.G. .,

The core exit temperature continues to increase due to loss of M.

~ .c; fc-flow to the steam generators and inability of the break to renove

, ., G. .' .

n-

.. sufficient decay beat. Figure 3 indicates that this is likely to I.)

1 .

" occur in the .short term fo'r breaks up to about 0.035 ft8 The 3 i.. ,

reactor vessel level which is established in the top plenum when i ,

yr.;.

' ,i ,.y a steam bubble forms contiiives to decrease due to fluid loss

• .?

through the break exceeding liquid makeup capability. The steam ,~.

bubble 'is at a saturation pressure corresponding to the 3 sn% - . - .

increasing care exit temperature. The steam bubble in the U-bend p.

y sf.-

is compressed and condensed by the increasing reactor pressure.  ?

• f} "

• I f

The D-bend is refilled with liquid and natural circulation should ,}

, I

%g he restored.- The restoration of natural circulation reduces core l w .g exit temperature and correspondlag top plenum pressure until a a.

g K A /

new steam bubble again forms in the lower pressure U-bend region t h,, ,

) x- .

& and the process repeats.

' cnce formed, the reactor vessel steam bubble should be sustained d k ' '

and grow larger to accommodate the net blowdown of fluid from the J. .

[L

.- system. ' The reactor vessel level which is established may .

.t

(*-

• )

{ ,

experience some small additional pertuhations as the steam I

"* bubbles form and co idense in the U-bends, but the short-tera . t.

i  ; .

trend should be for a decreasing level. A continuation of decay .

l ,

, ]- ..gq heat removal t,y intermittent natural circulation should be essured unti1*the U-bends can no longer be refilled with liquid.

During the natural,. circulation phase following a water-side .

.' . break, the pressuriser surge line and vessel slowly refill with sent-quiescent liquid. The pressurizer steam bubble is in ,

equilibrium with a slowly decreasing reactor vessel steam bubble

< 3; , . . . .

~.

. ~ ~ :: .;-

! b 37 sk+k.,

f O 12G ~,' 9.'j&

i l

l I _ - _ . _ _ _ _ _ _ _ . _ _ . _ _ _ _ . ~ . . _ . _ , _ _ _ _ _ . . _ . . . _ . _ _ _ _ _ , _ . _ ~ , _ .

J ..

...x

[~best capy V. -

'. - - n-++ -

= tu:.

-l. * . . /0 3

d  ;* ' -

,->v,.,. ..... j g .. ,,, ,

, , . [(. .

'j...'.

a

[4, ,

pressure and'is mostly likely slowly contracting as steam from I

..] the bubble condenses on the pressuriser walls during k:ooldown. - -

']

p 4 '

For a steam-side break such as at the press'urizer top, the break L q

vents the overpressure and flashes any saturated liquid in the ,

re .

pressuriser in a similar fashion to a water-side break / but the 4.

d level loss rate will be lower due to a lower mass flow rate for .: o 6

steam through the break (Figure 2). %ihen the pressuriser steam  ??

bubble reaches pressure equilibrium with the reactor vessel steam ..{.

bubble, it starts to contract rapidly as steam continues to be j?

.j I removed through the break. Fluid entering the pressurl=ce is at _

system saturation conditions, but some pressuri=er metal heat input remains to reduce the pressurizer level rise rate. One J

r

' entire steam bubble is renoved quickly through the steam-side  ;.

.g break and it becomes a water-side treak. The reactor vessel drain tim should be much shorter than given in rigure 5 because .

this figure is based on a constant pressuriser level which is a 1

5 valid assumption only if the steamside break is not at the L C . *

- pressurizar top. ,

.~t *:.. . . . . .m .

t 3.2 a'ransiti onJrom t:atural cir-ula11.gs,to Pool Pelling Natural circulation clearly ceases if the reactor vessel level .

reaches the top of the hot leg pipes and reactor. steam starts to ,

1 break away and bubble through the pipes and accur.ulate at the a

high points (which are the U-bends). Water in the reactor vessel p

J starts to boil agressively due to loss of circulating flow. ,.The reactor core exit terperature and corresponding top plenun stean I

bu'bble prc=sure increase. The hot leg pipe pressure remains J

essentially equalized with the cold leg pipe pressure by reactor' .

.e e .

g I!!3bM20@. - . foi2s-- " N '5E

, , -w-nw~.n,--n , - . ,. - - - - -~ m m,---~.w ---, , ,,c,.,,.,,- - , - , e,-m.,w,.,-w,,---,,- - _ ---,-..,,,w,---,, ---,-e,--en,-n

]best copy . . . . . = . .. ~

. . . . ~ . ~ . .

. . . a

~.

?.~n. ,-

- ~

5

// L; j

- - - .- . y.y n .u .

9 went valve actuation. The reactor coolant pump loop seals

/. 1.*

v ,*

6

- , 7.,:

T El #

'. inhibit reactor steam entry into the steam generators through the i.

"v cold leg pipi.ng. The water level in each vertical het leg pipe ,,

e decreases due to fluid loss through the break exceeding liquid c ;.

. W O

j makeup capability. .

- since natural circulation is no longer possible, the core must.be .

. coo. led, in part, by pool boiling in the reactor vessel with -

D . .r acadensation inside th's steam generator tubes ar.d pool boiling on @

i 8

c. . ' .' . The steam generators are assumed to be ,$'3, F; 9,.e:.secondary side. *

' isolated and pres,surized on the secondary side to the.1owest safety valve' set point. The initial secondary side liquid Ie, vel 4

is deter,mine'd by plant operating conditions at the time of reactor trip. It will most likely be in the range of 6 to 26 ft j above the bottom tubesheet plus any net addition from auxiliary J

feedwater.

The transition from natural circulation to pool boiling may be j

troublesome because of the time delay incurred while waiting for ,.

the water level in.the U-bend region of each hot leg pipe and in--.

~

the steam generator tubes to drain below the secondary side water

• 1 eve l. During this ' time, no appreciable heat is removed by the

. steam generators. The steam generator drain time is given in By definition, this drain time starts after natural rigure 6.

circulation is lost. The drain eine represents the minimum time during which system repressurization will occur if all decay heat is not being removed through the break. ..

1 .

t

a. ' .;.

l 012G '"'.' J U U

. I l!3hM-7.2 .

y e

- - . - - . .,,,e-. - . - . . - - - - - , . - - - - . . , , . _ , - - - ,,---,----s.----r,,-.

Lbest copy J .. .. =- 1. .

. _ , g.

mL - . ..

^

/*"2 }

i

,) .. -

( ,

. I* .

~

. . c,.,. -? w ,.~,. g q... g. ,g , , ,

Figure 5 indicates how long it.will take for. the reactor vessel

• ' .'.l. .

t 1 top plenum to drain after a tireak. The actual higher pressure

• T'

.N'lr

. . .c .-

-).; during a portion of *the drain phase will decreare the drain, time. ..

(

Fioure 4 shows how Jong after a break before all decay heat can T,f f . -

j be removed through the break. For the range of smaller breaks, [ ,_-

g l reactor vossal drain time for a given break size is somewhat f,.

,.T shorter than the energy equilibrium time. This means that R. '

h .

natvral circulation ceases before the break can remove all decay . .e

--("4 heat. system repressurizatlon will occur as required to remove .

the' excess decay heat .

while waiting for the steam generators to g, c 2...

-g drain following the loss of natural circulation. Incress'ing the f

y a -

jr pressure increases flow through the break and thereby decrease's energy equilibriun time for a given break size. 2t should be u

• noted from Figures 4 and 5 that repressurization to 2500 psia '

1 appears unlikely since all other curves are to the right of the

.j 2500 psia. curve for c3 = 1.0 which is the bounding condition.

f Although the results indicate that full repressurization is i

1 t'nlikely, it should,be understood that the effects of partial 4

repressurization have not been evaluated in terms of minimum core level and peak clad temperature ef fects. The calculations which were performed merely confirm that repressurization may occur in order to remove decay heat when natural circulation is lost. ,

Increasing system pressure -increases flow through the break and .

1 decreases makeup available f rom the high pressure injection pur:p.

I Tais will probably result in a lower ultimate core level and a

\.

. higher peak clad temperature if the core is uncovered. An Eccs g

I t

\ type analysis basei on an t.ppropriate model for the very sna11

~

I becak LOCA is required to determine the numbers. .

S .

[ O126. k E.' @.

.....-...s ~

.  ! II 3l'.k.20N..I

--..-,-,n. - - - . - - - - -- , , . . .,._--.-,----.,.....,-.,--w- ,-,-_.,,--- ,,,-..- -.. .--, n, , , , , , . , - , - , , - . - - - , - - . , , . . ,

(

Loest cupy J. -

- . r ., . .

)..' .

w, l

l *, .

. . /3 ,

y ,..

.., v .. :. . .

2 ... c ,.g . n. ..=.p.,.,..,,.... .

g - i, 4

,During the transition to pool boiling following a water side 9, 3

.M

• • 1 1

hreak, the pressurizer surge line loop seal inhibits steam entry sc-A. -

O

! nto the pressurizer. The pressuriser slowly fil.ls as ths !3 J. ,i - r e. :

remaining steam bubble in the pressurizer' is compressed to system T;, - < .

pressure and condenses. The water' 1evel in the vertical hot leg 3'

') s

e;..'

3 .

pipe eventually drops below the surge line connection to the pipe ,

1 , -

o -

y (Figure 1). Any further increase in pressurizer level comes from

' "f l-

4

.- water remaining in the loop seal. The loop seal is soon purged

.y O.,-

l 4 '

' - and steam from the vertical ho1! leg pipe passes to the j A. .-

? t rj ,

.M pressurizer void space as required to compensate for condensation .'

i .

( ,I of the steam bubble. The level should stabill:e. ,;

l ,

i For a steam side break such as at the pressurizer tep, the pressurizer may refill before the reactor vessel top plenum is . y 5 --

i drair.ed and natural circulation ceases. It should remain fille!

k unless the vertical hot leg pipe drains to below the surge line connection. In this event, water flow through the surge line to the break changes to steam flow and water in the pressurizer and ,

.l y

surge 'line is heated to saturation temperature and purged from

~

the system. steam bubbling through the water to reach the break

. .h. .

may create hydraulic instabilities.  ;"' .

3.3 Pool EM11gg .

e

} h.*'p C In order to condense steam inside the. steam generator tubes, it i is necessary to drop the prirury side water level inside the

• . .E y'\

k!j' tubes to sonewhat below the existing secondary side water level. -

- n.

.t j.f . 'since het and.. cold leg pressures are virtually equalized by' ..

W reactor vent valve actuation (ninimum AP of 0.15 psi to open),

the primary side water level .:-

is the same as in the vertical bot -

t iilsj@.?@ ,

o12s JY . @+- >

e.

- . , _ . . , _ , . . _ - , . , _ . - - . - - - - - - - - ,,_,__..,,..-,__.-__-,_,,,..,-..---._,.m,,.,,,,..,,_ .-w-,--w--_-,,,, , , , , , - , . . , , . . - - , _

I L Dest . copy s.. . : .r: .: . . . . . .. ..

r  :.

4 M [--

g., .'c q.* : e

) *

[ "T**.** " ' E P h . d '; i. * *j ;, , . . . , .,-. , , . ,,...,,. ,

• f jy,

• l .

leg pipes,(except for a density diff erence correction). A g 1

1 p ,b..,..! .

portion of the steam which is generated within the reactor core l.;,j;q.6

{

will seek its way into the vertical hot legs and undergo bubble ((*,.

I.

I . :r. s

• disengagement at the water-steam interf ace. This steam is then 9,

~f '7

[ frae to condense inside that pcrtion of the steam generator tubes 2.0 which is above ~the primary side water level and below the r. '

l s

e A *. .

i secondary side water level. Condensate inside the steam generator tubes'is returned to the reactor vessel by gravity flow $

Qk*-

jg.

through the pump loop seals and cold leg pipes.

. CdI.

p.'.

j The reactor vessel level (two-phase) fluctuates above an

, S' elevation corresponding to the top of the horisontal hot leg j i).

) pipes until the. level in the vertical hot leg pipes drops to the horise'ntal hot leg elevation. At that tine the reactor vessel {

l J 1evel extends into the hot legs. It then decreases until any - i

( i e

/* fluid lost through the break no' longer exceeds the liquid r.aheep

• g d capability. ! For certain small break LC M s, the reactor vessel

?

\,V le turnaround may not be reached until the upper portion of ,

-j .

s re has been uncovered for a prolonged time. For certain

. D ler breaks, turnaround might be reached during the natural ig 4circulation phase or transition from natural circuation to pool *

\.

i~.

%.x ve a boiling.

j 3.4 IIin1111g2 fro

• P291.!!olli"e t9 E112ral circulation -

4 s

.Following reactor vessel level turnaround, the level starts to 7

. increase and eventually returns to the top of the horizontal hot - ' '

leg pipes. .A water level then appears in the vertical hot leg

[

f 1

pipes. A corresponding level (execpt for a density difference correction) appears in the steam generator tubes due to pressure

. .e s 9

' c

,liab5FdE.6i *

• . . ;n-we %, ,

012G "

l

o t '

copy 9 - -

..-- Lbest .

..-e. . _ .. _ ..m .. _ ,.n;

/.C"

- - ,,. . ..  ?. >.-

• W.v: '; ':

y .**r, .,.

s.,... , . .:. g , . . . 7 4 equilization by the reactor vent valves. Any further' net .[.-

.g

.1 addition of liquid goes into filling the vertical hot leg pipes jj,r

, h , and steam generator tubes. A portion of the steam which is '[

generated by. decay heat within the reactor core seeks its way - ..

^-

y, l into the **ertical hot legs, disengages at the water-steam . . . .;

a 1

interface, and condenses insiJe the cooled region of the unfilled portion of the steam generator tubes. ,-

~ll

• -- I-h .

[8]

The reactor vessel level fluctuates above an elevation . g 1 corresponding to the top of the horizontal hot legs. The level

.y should slowly increase as the large steam bubble in the upper  ?

1 plenum is condensed under the influence of the hydrostatic head ,

produced by the rising water level inside the vertical hot les I

}

3 P ipes. ,

l 4

~

~4 j Decay heat renoval is accomplished, in part, by condensation .

Inside the steam generator tubes

• 11 the prir.ary side water level

-* is sufficiently below the secondary side water level and the accumulation of any noncondensible gases in the steam generator ,

9 ~

tubes does not inhibit adequate conSensation. Decay heat removal 'I t

.- by condensation ceases when the water level inside the steam

~

. genert. tor tubes becomes greater than the, secondary side water level. If the break cannot remove all decay heat, water in the ,

reactor vessel starts to boil more agressivsly due to loss of one i .,

of the required heat sinks. The reactor core exit temperature

< .\ D and corresponding vapor pressure increase until the reactor vessel top plenum steam bubble becones the conti'olling high h pressure. Makeup water continues to fill the vertical hot leg pipes and steam generator tehen. The steam bubbic which is trapped inside the U-bend pipe ab'ove each steam generator is 4 IIlf p M D .. ..

i o12sl3 O O l .

.a ~Lbest copy .

j ._.-ei"...,... P .

. . ..... .? .

' * . . . lf l  ; .

5, . ,

...-c.,..

8 .W. = . ,

, ,, y .

< c

., slowly compressed and condensed by the higher reactor vessel 0

%i -

'y

. pressure. However, the D-bend may contain noncondensible gases .y

?

-A tonich have accumulated at this system high point. If sufficient e / /

l , '/ noncondensibke gases are present, it will be impossible to refill ,

',i '

/ the U-bend witt fluid and establish natural circulation. '.

E '. %

.1 l If natural circulation is re-established,'the, reactor vessel .N

l 1evel is above the horizontal hot 1*g pipes and increasing, and

. .f, j

the reactor coolant

• piping and staan generators are full of h.-r.. -

]

water. The reactor vessel level continues to increase until /.

!! operator action is taken to trim back on h makeup rate. If ,

Qs

-, natural circulation is not re-established and the break cannot remove sufficient decay heat, the reactor coolant system pressure J

I

• increases until adequate heat removal .:an be achieved through the ,

break or the press'uri:er safety valves open.

j.

At some point, operator action may be invoked to open the .

e pressurizer electromatic relief valve to assure continuation of

- the more stable mode of pool boiling for decay beat removal or -

i . . .

- provide suffic'ient depressurization to go on shutdown cooling.

' Nowever, this' valve has not been qualified (Class TE power, etc.)

L to perform an essential mitigating function and can be

~

inactivated by a postulated single failure.

During the transition from pool. boiling to natural circulation, t

3 the vertical hot leg pipe starts to refill and cover the surge 4

line connection to the pressurizer. Th; pressurizer completely .

as the trapped fills (except for noncondensible gases) g pecesurizer stean butbic condenses or is vented tlacugh a steam-g , space break. .' " ~

~

we

!L3 b W -i-@.'.F . Ioi2s '. ' ' ' . ads

\

l . ,-

__-..__-.,--__.m._ . . . . - . _ - - _ . _ _ _ . _ ._- _.___- _--___ _.-_...____ ___ _____ _, -

l l

l .

i l

l ' ; . ?.:. :__ ' . ; -

cepy . 6.f;_..f.Y , , , . .- I Qbest i

, .- /7 ,

~

..'. ...s..

..y..r f

's. .-

3.5 shutdown cooline .- --

(".t,. -

P .%, .

4 In the long term, react::r coolant system pressure will be reduced to below 3$0 psia by blowdown effects of the postulated break.

• A

. ,5

- If the horizontal hot leg pipes are full of water at that time, y it should be possible to remove all decay heat through operation *

w. -

of the Decay Meat Removal (DMR) syste:n. There appears to be e3 adequate available net positive suction hea6 (NPSH) at each DER ,

pump suction to assure acceptable operation at saturation

, 9 "i

- ' canditions if the flow rate is kept low.. Any fluid still being ,. ,

1 . lost through a breas can be made up by one of the DnR ,um, too,..  :-

\ u.. a .

taking suction from the r *"ni .5 water s,torage tank or by a high y pressure in[ection pump loop if a postulated single failure 1

- involves ene of the DMR loops.

I J

' *js 4.0 WOR 37 CASE 3.DCA CQt4SIDERATIONS .

After identifying certain characteristics of a very small break .

,.1 toCA and the various modes of post-IccA decay heat removal, some g.

1 thought was given to identification of the probable worst case ,., ,

f' /

' for safety analysis purposes. A nt::bar of considerations were -

$ investigated briefly to evaluate their likely influence on the 4 worst care seicetion. The more important of these are detailed

- - ...a kelow. ' *

=.. .

4.1 ole.chirne m effleient ano ar P in atte . . . , ;; . . . , . , ,

.s-  ! -- -

. ,. ..~. .

i ~j .

Figure 5 indicates a Moody coefficient of 1.0 'i nstead.of 0.6 for .

water-side breaks shorten = significantly the reactor vessel top

'l plenum drain time for a gisen break sisc but rigure a shows a compenesting reduction in energy equilibrium time also occure due N

l

"* r.

D.W- /)q%

3 - ,

~ 'r' .v 'pfV u 3i . . i .o

, l 012G r' .

e

,m,. , 7, ,_ , , . _ _ - - _ _ --

.m..- ...,._,.,, _..- _ , ,----, , ,,,_._,, __.w. . - - - - - - - . - - - - -. _ . - - - - - _ _ . ._ __

. h: .

_ _.. _ _ L best ' copy . . . . . . .

n.

. :v. ..

q,

/$ .; .

. s ., .

. ,. ,..s . .-

~ p;

.9 ., ..

3E N

} to increased mass flow. The net eff ect is that only for water- g..,.

y .,

' side breaks less than 0.02 f t* with CD = 1.0 will reactor drain - R.;.y s time for a given break size be somewhat shorter than energy equilibrium time. For this case, natural circulation ceases $. .'i

• h-j before the break can remove all decay heat and some system g

. '? .

repressurization occurs. For a water-side break with eg = 0.6

. 'I the comparable situation develops for breaks which are less than u :.

0.035 fta. In both cases, the dif ference between drain time and L

x,i .. .=.

F-Ti 2 energy equilibrium time is about the same. From the viewpoint of '

reactor vessel drain time and energy equilibrium time for a given  ;

break area, the conservative choice byer the range examined ,-

'\, appears to be,g =0.6. ,

Figure 6 indicates that a Moody coef ficient of 1.0 instead of C.5 for water-side breaks decreases the steam generater , drain tir.e ,

for a given break size. This assures an' earlier uncevery of a ,

l 4 . s cor.densing surface inside the steam generator tubes; therefore, the conservative choice over the range examined again appears to be. en =c'.s..

t

. _. .i .. -

In applying Figures 3 and 4 to a specific break, it should be determined that the fluid lost through the break remains representative of the fluid at the core exit. An arrangenent for adequate mass transport twater, steam, or two-phase) from the

, core exit to the break location 'must be assumed if the decay heat For certain water-

.. . . is to be removed ef fectively by the break.

l side break locations, the high pressure injection (HPI) pump flow may bypass the core and any d. cay hett generated within the core .

say not of fcesively,co:.nunicate with the subnerged break or steam ,

. .A '

g s

4 _

i.. ., . . . - . .

. ji 3j $ , Q _.E i. 012 6 [. r' y>>7'

. _ _ , ,. ._._.-_..,__m . _ _ ,_.m ...,__,,,_..,-..,_,.,,.,.-..m,.,.,.._,,,_,,.--.,_.m _ , . _ , , . . , , , _ _ _ _ . , , , , . . , , , _ . .

.s .

f t best.

. , Q. .

C0py * $ .,. . n * . ,.; ,

_. 9 f: 19 gj

. ?,

- . . ..,..n.....

. f.

generator tubes. There may be no significant decay heat removal  :?

. . .-l. .

while this condition persists.  ! , f.! l

.: j 4.2 gg g gat Demoval 1

N Decay heat can be removed from the reactor coolant system by 4

i blowdown through the postulated break and by a release of steam m.

[i,

- on the steam generator secondary side. The break is effective

.l ,

't

' for heat removal at all times unless it becomes isolated. di be .-- . N. . -

-f., a l steam generators are effective only during natural circulation or ,

A 1

-8 Ji :a after the primary side water has drained to below the secendary side water level and concensation has become ef fective. Natural 3

.g circulation.is , lost af ter the reactor vessel top plenum has j The steam generators be:ome effective again l

6

]:. drained (rigore 5). .

I g after the tubes are drained sufficiently (Figure 6). The steam y i '

\ generators are no longer required af ter all decay heat can be 8 removed through the break trigure 4). ,

e The various modes of post-LOCA decay heat re:noval discussed in ,

section 3.0 occur unless level turnaraound develops before the 5 3 -

pool boiling stage. Most water-side breaks which can be classified as a IDcA lose natural circulatica and reach the pool l ]

boiling stage before level turnaround. Many of these breaks ,

reach energy equilibrium through the break with perhaps some

• prolonged repressurization before the steam generator can drain i sufficiently to become a condenser following the loss of natural

-)

circulation. As a result, completten of the reactor vessel top - .

m

\ plenum daainage through the tren (which cuir.inates in a loss of '

}

• natural circulation) appears to mark the end of any essential

, usef ulness of the st,eam generators for very s:.411 break Loc 1.

.-e ,

,"'"O Y l

'U3hh... .%.D._d s . . { O126 l

t

• l l

.' Lbest hopy -

m. . 4 ,

j .

f u

a 's ' .:y' ~' . u . .z. , .,s .

. ~ . . , .

3. . , .

.J mitigation. Note that the natural circulation phase appears

• A s ,,4

. '.ty. .,<. ' t.

.. .v, .  ;

,,,,,gg g, .

4.3 f.evel Titrnaround agj EEerev revihibrium l . .  %

.4 1mvel turnaround occurs when makeup available from the high

• i pressure injection (MPI) pump escoeds mass flow rate through the

break at the existing system pressure. Figure 2 shows that at 2500 psia the flow from one SPI P88P exceeds the mass flow rate

( _ '.

through a 0.004 fta' water-side break (C3 = 1.0) 'or a O'.008 fte

'i l'

'J

- ,' steam-side break. 14 vel turnaround should be immediate for these, N break sizes. f the break is larger but less than 4 01 fte for a water-side break or 0.035 fte for a stoa:n-side break, level 1 -

a s re., reaches 1:70 psia.

j turnaround occurs before the syste 3reaks exceeding this range lead to a prol'onged loss of fluid Q

' -D* .~ through the break with p'ool boiling at 1270 psia until all decay heat can 'be removed through the break (energy equilibrium time) l g-N and thereby further depressurize the system. steam generators 1

4 Mv f' / -

cannot depressurise the primary system to below the set point of i

i 4 Y

• the secondary side safety valve unless operator. action is invoked t' n/

i t

l to open atmospheric dump valves. The delaying effects of reverse

1. heat transfer frca the steam girherators must be accounted for when depressurizing through a break at below 1270 psia. Figure 4 shows the energy equilibrium time for a 0.01 f te water-side break with C 3 = 1.0 or a 0.02 f te water-side break with CD = 0.6 is over 30 minutes.' A 0.035 f te steam-side break' reaches' ,;

equilibrium in about 5 minutes. Natural circulation must be ,

a ,. , , assured yhile awaiting encrgy equilibrica if repressuri:stion is J

y to be avoided.

1 1 .

' hsj .h5'~.@ i. . o12e 9 J'6d

. O y e

, - - - . - , , - - - - , - - - , - - . .. ,-m,--, ----.--,-.----.,-.,---e-r-- m._,-~,,-..---,-- ,,,_~,,,-.,-,,-e,,-.m..,,--,--y. . . , ~ , . - .

. l

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

i best copy ( h.. . t , .: .. . .v.

- . u Ji 3 l -

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

. . . , - . g . v ..,,: .,,,.

_ . . .. ...w. pp .. . . . .; ., . . . _ _

It is assumed that natural circulation ca'n be, maintained until ,g,.

o .c d,Q

.'2 the reactor vessel top plenum is drained. Figure 5 shows reactor .

4..- .

• vessel drain time is about equal to energy' equilibrium time for :v3 y

1 Therefore, when natural ' ' y, 8.01 ftt water-side break with CD = 1.0.

h circulation is lost the break should be able to remove all decay ' (("I V

]l :

b4at with no fdtther need for the steam generatos as a heat sink. -

i t

s. .

For water-sida breaks in the range of 0.01 - 0.02 fts with ,.

j

• a T.,

CD

= 1.0 the reactor vessel top plenum drains and natural 'q . s .

eirculation is lost up to 5 minutes before energy equilibrium is .

f achieved. Figure E shows the steam generator drain time to be .

t I

from 10 to over 30 m'inutes. Therefore, energy equilibrium is, . ,

i < astablished before the steam generator becomes an effective heat 4 -

I sink. Some system repressurization vill occur during the' 5 h minute delay. For breaks larger than 0.02 ft8 with Cp.= 1.0, t

t energy equilibrium is reached before the reactor vessel top ,

plenum drain is completed.

i -C 1 A similar situation exists f or water-side breaks in the range of ,

A j -

0.02 - 0.035 fts with CD = 0. 6. The delay times are about the. ~

I same. For breaks larger than 0.035 f ts with cp = 0.6, energy

. I equilibrium is reached before the reactor vessel top plenum drain is completed. ' .

For all steam side breaks up to 0.05 f t*, the energy,equilibrius

- V time given in Figure e is always much less than the reactor

! vesseltopYlenusdraintimegiveninFigure5.,. In every case, ~ -

> ~

j

' the break should be able to remove all decay heat well before natural circulation isi lost. ..

4 . . , . ,. ..

1 1 .

. o' p'.m, 0126 P' 1 s '^-

Is..:,., :. ;'.9. ..,. , .. .c. 4 1 -

. ~- - - . -

e

--~.------+--.----_.-..-...,_.,,_...-_,,.% _.m- - . . . ,m,,,-%_,,-,,-,m-,,,,--,.,,__,____._.,,,m_,m.,.,#,_,,,,,.._w ,.m _,.w-b.,.,,w..

.";.. . . . .,. L Dest copy 1. ... c. ., ..

• x' ..

. .) . ....

..y .

. ... + .. : .

..u . .

Many postulated breaks change from a water-side break to a steam-  : ..

l -

l

~ side break (or the converse) sometime during the accident  ?.i .-

. %.% scenario. For the case of a water-side break changing to a

.  ?,-

J . V steam-side break, Figure 5 indicates the reactor drain time 1

! increases considerably if the change occurs while the top plenum 1

- is still draining. Figure 4 shows the decreased mass flow rate through the break when steam starts to flow does not have a .

r. -

'a detectable effect on the energy equilibrium time if the previous ...

Q.J

.; enter flow is for CD = 0.6. If the water flow is for CD = 1. 0, , {s the' energy equ.811brium time after steam starts to flow increases h

" noticably; however, a comparison of Figures a and 5 indicates .

• 3, ereggy equilibrium time during the steam flow phase is always

very shor. 'when compared to reactor vessel drain time so no

,3 e *

• zepressurization effect is anticipated.

4 I If the postulated break changes from a stean-side break to a  !

water-side kreak during the accident scenario, rigure 5 shows reactor vessel drain time decreases markedly. The break o .

i characterirties become those associated with a water-side break j and some system repressuridation may occur. This is the type' of ,

• r

'.i' accident sequence which develops during a break at the pressurizer top. It initially vents' the steam overpressure and everstual'ly passes water. * - *

.'M -

1 - -

4.4 Ksss vendE r calculd11232 .

I <

[

q , ,,,,

A 0.05-f ta break at the pump discharge is the s'aallest break . ,

.. . analyzed and repcrted by 5 & W using their 1:RC apptcved ECCS ,

evaluation model. Their resuits indicate that one HPI pump alone is sufficient to handle a break of this size. Although the

',' ' 0. ,p.

'O -

! O126

... ~

'.3

. I!2.1:.2. ;. ; :~.&96f.

w

, , -- w - - . - , - w - -,

,,-n- - - , . . .a,-.--..w-,-- ,..--g-w -

,,a.=,e n--n-- ,,-.-w.--. . - . . . .m- - - ....~.s..-,s.,--- -=a,ewme

L.best copy J. .. _

E -

7

)+ .

,* +.,

. . .. p3 's .

. ... . . - - .~ . .

seactor ves'sel liquid' volume is shown to drop to the core top

.j

{ .i.

t..- .

within about to minutes following the break, the results confirm V,f

.tn -

that the core remains virtually covered. pith two-phase fluid at

.:c .- "c . '

t all times and the fuel cladding temperature never exceeds its

. 2

< prebreak value. }*)

i Figure 2 indicates the water flow from a 0.05 f ta break at the ,..

.. . o g pump discharge is considerably , greater than the capacity of one -

. 1 l EPI pump at 1270 psia. Figure 4 shows that energy equilibrium f. g .

}( through the break is reached within two minutes and Figure $ 'N O

4

(

1. } shows the reactor vessel top plenum drains in about the same time. Therefore, the steam generator does not have to function [.

I g i as a heat sink beyond this point. Depressurization below 1270 g V .

~

A psia becor.es possible af ter energy equilibrium is reached, but .

the rate will be slow because the steam generator heat surt be ,

i ' removed by reverse heat transfer. During this time, the reactor coolant system continues to drain through tho' break. 1.evel turnaround most await a lower pressure and commensurate increase .

~

in pump flow. '.

It should be recognized e.lat a 0.05 f ta break'is .near" the lower size limit for the rect

• valuation r.odel and near the upper limit for a very small break LocA analysis. The ces evaluation model ..

does not appear to take into consideration the possibility of .;

'.l NO

! inter-f.etent natural circulation' or the ef fects of s,' team

d g.c. C generator drain time during the transition from natural, l

N . circulation to pool. boiling These ef fects are important for the ,

case of a very small break LccA and will be experienced before

-" the reactor vessci level reaches the core top. They may not be .

1 . ...~

' considered important for an Ecc3 type analysis of a 0.05 f te

!.H4:.5W.$,5! .

[ o12s N O 57-i

, - . , , -, -------.---..--,,,,,.n.. -

- , , _ .. ,-.., ,-... -.. -.._., - --- ,-.---- ,-e- - - ,,. , , , , _ . - ,

l copy

c. - 'L b est J .. ,. .

- x.. .

5-l M ,

* *.* . . ~

I .

. .j. break because the decay.haat rate is in equilibrium with the heat .&

(.,

/a i

' '?

lo'st through the break in less than 5 minutes following the break ,.

(Figure 3), thereaf ter eliminating the steam generator as a Y

.* r required heat sink. For smaller breaks, the loss of this heat  ! .-

]

sink and its ultimate ef fect on fuel cladding tenperature needs to be considered. .

h a

H i 4.5 Break Tsolation aD LPo-o shutoff Effr.g33 N

.n - It may be a natural ten 62ncy for the plant operator to isolate a - 2g; t

very small break if it c.an be located and valved 'out. This may T l

even be a requirement of the emergency operating proce*are or plan. In some cases, such as for a letdown line, the isolation '

maybeautc[atic. Break isolation may be partial or co=plete as

.a I deternined by locatica of the isolating device, number ant site -

4 j of f'.cw paths to the break, and possible variation of effective

$ break area by opening valves such as the pressurizer vest valve.

complete isolation reduces the break area to zero. The rer.alaing ,

4

~

'; water inventory is determined by the original break area and the tive aft'.er break before isolation is achieved. For instance, if '

7-u '

the original break area is 0.05 fte and CD = 1.0, energy i

equilibrium and loss of natural circulation occur in less than 2

$ sdnutes (Tigures 4 and 5). At this time, the break can remove all decay heat, but the reactor coolant systen continues to drain (Figuce 2). Depressurisation below 1270 psia starts but is impeded by reverse heat transfer from the steam generaters. At,8- ,

minutes af ter the break, the steam generator tubes are drained down to the secondary side water level. If at this tire the l ,

break should be isolated, there is no etfcctive heat sink (water

.  ?

2 - - -

(l3.1.E#09,'S

.i o12s [" "M7-e

,- ,--,,,,,--,,--,,--.,w.. ,.- ,n- -e,,,n----,,- _- .,..--.n, ,.,-----_----n..,.,,.--n, ~.- -,-,

1

- T-y .._ _..e .

. . . - Lbest c'opy . .. .

--... . :.h . : ,._ .- ...

, m_, .

y v p ' ** * *

• *  ;:,y

,.

• a*'**

. ... . , . . , - . . ,.* :-- + .'.. Y . - . v. .. .x;l.1 I

a.-

cooled condensing surface not established in staan generator). M.*

'tL* .

1 The systes starts to repressurize and refill. Nowever, it takes lj[.

9.-,';1 1

at least 25 minutes at 1270 psia to refill the steam generators ,

j. I and re-establish natural circulation (ylgure 7 for Cg =1.0"and 2f, , ,

Y'  % ..

ft initial level). Repressurisation to 't500 psLa appears likely ~!

, with a commensurate reduction in makeup flow and eventual opening g. ,

1

• of .the . . pressuriser

. , . .safety

.. . valves. as required to remove the decay.

1 ;1.*

.--- bean; <' y.,

14

. .. 'r. -

J c *

=

y.M.g .; ..

d The systes now behaves as discussed in'section 4.3 '(steam-side .T,.}e P. '

?

.)4 - break changes to a water-side break) as any rer.aini.sg pressurizer

• bubble is vented through the safety valves and the pressuriter is O I

filled with water from the vertical hat leg pipe. The impact and .

passing of water through the safety valves may create hydraulic 2

} instabilities and cther service conditions for which the valves l have not been qualifiad.

I A rapid filling of the pressurizer free space with liquid.

produces a corresponding level drawdown in the steam generator b -

tubes which then exposes a condensing surface. When sufficient >

]' surface is exposed, the safety valves no longer need to open and l the steam generator tubes start to refill. After a tisie, the -

, condensing surface is flooded again and the safety valves reopen l , to rer.ove the decay heat. The alternating renoval of decay beat f through safety valves and by condensing inside steam

• generator I

l

• l . tubes continues at,2500 psia. The reactor core should remain r .

l covered. During this time the reactor vessel steam bubble i

g controls system pre:sure and suprorts the vertical hot leg water 3

colurn and keeps the pressuriser full of saturated liquid.

)l - . . ..

1 ,

~

I '- .__ . M 1

U ' I'

{R-i.R2.',7.~L . . [ ' 0 12 6.i _

0'-

- , , , . _ . - - , . _ . . - , - . , , ,e. .m.--,--.- -

bbest Copy -.. . -

4 ".

- . _ .. . .I .

• J (,

• *****' c .. a. e. . . ,, ,. . , . , , .

.m smaller area breaks which are isolated shoeld behave in a similar '.),.

,:.:k A., IP

.ashion escept it may take a longer time to lose natural sirculation. , larger area breaks can 'also produce sinitar

- 2; 11 circunstances if ,they are isolated. , c g.

-t.  ;

A full pressurizer may convince the operator to trip the PFI pump

% an / watch for a subsequent loss of level. If this happens cad ,.

the break hAs bien isolated, "the sda generator, tube level starts -

" to decrease due to release of fluid through the safety valves .

)

) l until an adequate condensing surface la established. No further

? 1 level loss'is likely and the safety valves should remain closed.

'i The pressuriser should'

9) [ A stable boiling mode will prevail. _

ree.ain full of fluid with a controlling steam bubble in the y reactor vessel.

4.6 F.Igr,13r,iggr ?.evel Yng.ication f

." The modes of decay heat removal discussed in section 3.5 point

., ~

i \ out that pressuriser level is not a correct indicator of water- .

f .,

r.

,1 \

level .over the reactor core. During the natural circulation H.

g .

• ' phase, water can be draining fron the reactor vessel top plenum

'h while pressurizer level is slowly increasing. If the break is at

.:, the top of the pressuriser steam space, a rapid pressurizer r o refilling can occur. During the transition to pool boiling and while in psol boiling, the le4el should stabilize even thcugh the .

core may be uncovered. .Therefore, precsurizer l'evel is not

- . . - . . v. . ..

considered a reliable guide as to core coollr.g conditions. No. ,

other primary side level indication is psevidad. There is a full range level indicator on the secondary side of each steam ..-

~

., j generator.*4.= .

s - . .

• • N _

l 012s V90E I

. IilshiGL9[.N .~ .

men ud

-l I Lbest capy .. ....

).

i -

.p* */  ?

-) . . . o

• A .

e .. - ., . ,

(. =.~ r , . .... p p.

w- * . .

c4 , . .

3e ,

j -

A simLlar problem with pressurizer level indication is found in i  : .

T.g

. ..A ~ . A full pressuriser may gf.

section 4.5 relative to MPI pump trip. <. .

,  %.r convince the operator to trip the MPI pump and *O watch for a

~

f.

Although this response appears d

subsequent loss of level.

desirable, a full pressuriser may not always be a good indication

.,,j For instance. . h I

of high water level in the reactor coolant system. >;

t the steam bubble which is trapped in the pressuriter may be -,

2 * .

,& vented by actuation of the pressuriser vent vslve due to high .

ge s..:

pressure developed in the reactor vessel top plena or by

.y],

4 The vent valve will subsequently close but the operator action. The. ./

gessurizer may be -filled solid with a subcooled liquid.

aoop seal configuration of the pressuriser surge line allows.the 3

pressuriser to remain filled as the reactor coolant system water -

level drops until system pressure is below saturation pressu=e .

• of.

s f 1 This may take a long time if the pressurise'r liquid inventory.

system pressure is set by a requirement to remove some of the Thus a full decay heat through the steam generator at 12'f 0 psia. Y jeessurize'r' is not considered a reliable indication for prescribing -ertain operator actions such as MPI pump trip. .

/ .

5.0 CONct.U520Ns .

l -i .

j l The results of this investigat'.on verif y the pressence of a class f of very saa'11 break LOCA's (probably $

0.05 f te) for a a C W 205-yuel-Assembly PWR 9hich may experience one or more

• impediments to decay heat removal which need to be better ' -

.f understooi if proper operator response and adequate sitigation The following nitussions have been identified

'[ are to be assured.

as special items of concern which require confirmation using

.../....... . .

. I 11n .RE2. 9 i o12c ' 4.'30 '! .

. ~ _ _ _

-..-..._,..-_.-.-.y ., .-. ,,.-,,_y, . - _ , _ - . . _ . _ . _ ,_, . ~.., - __ - e - --, ,.----.m_

1

)

. '. , . :.w  ;;,-. ~ . .[best copy ] -.- - - -

2. :,1

• . . JT .

'.~

. s .  % * .. . . . . .

m .

? ;.

.j

• detailed transient calculations based on appropraite system and '%

core therr.al-hydraulic models. The reported NSSS vendor models do not appear to accommodate these very small break IcCA .

I situations. .In each case, fuel peak clad temperature is the i

l parameter of particular interest for comparison with the ECCS .

I l

acceptance critieria, but stability of the fluid process and .

. c.j

. adequacy of instzunentation and components should also be

• E'-

eensidered. . . . *

J,i.U,52 '

'; i:

.r Intermittent natural circulation is identified as a possible) g.I 1

1

) 4 mode of initial decay heat removal following a very small

.I

' ( break 10CA (section 3.1) . The adequacy of this unstable' mode

, 4 for decay heat removal needs to be verified. '

g

=,

l ~2. The transition from natural circulation to pool .

g boiling / condensing involves a time delay incurred while waiting for water incide the staa:a generator tubes to drain Y below the secondary side water-level (f.ection 3.2). During 1

4 tihis time, system repressurization will occur if all decay

~ ~

E beat is not being removed through the break. The ef fect and 2

..\/ acceptability of this repressurization needs to be determined. . ,.

' s

3. The decay heat fraction which is renoved through the break

.for a gf ven mass flow rate will be less than predicted unless the fluid enthalpy upstream of the break is representative of the core exit enthalpy (section 4.1) . The sensitivity to .

- . upstream enthalpy, particularly with regard to system

.f

- repressurization, needs to be evaluated for those break .

l s .

locations wherein come core bypass may be possible. .

7 I

i _

~ '

" d'M

' U sh5iW'.000:

l . . o 1 2 s__. _

9

• w

.--gwwsr-v,-,ww-m-----w-w-

e I

o Lhest copy '

. . m' . -

".e.: ._

___._.,c.__._ .

- . w. _

-s

.,4 .

39 H .

,..e. . . c - . . ~ e . .. . . . , w . . , , .. .,,, ,,, ,

.. e ,

, . '.i

t. ,.The pressuriser level indication is not a correct indication ca 1

l .) of water level relative to the reactor core (section a.s).

) 'The safet.y significance of this shosetcoming needs to be .

== evaluated with regard to adequacy of inforr.ation for .- .

! +v, .

1 A*

corrective operator actions. .-  ;,f

.,=,

5. The possibility of.very snail break isolation by *s m eer 2 -

4 action and the subsequent loss of both the steam generators J.

- and break as heat sinks is of special concern (sertion 4.5). p' .

s-

, .6)

^

s The rapid repressurization' and' eventual exposure of the j,-l

'] -

f pressuriser safety valves to slug or two phase flow .

needs ~

m further analytical consideration and possible test .

qualification of the valves.

3 i

6. There may be a potential for serious process disruption or unacceptable functional or pressure boundary damage to co ponents and stcan generator tubes due to the hydraulic 1

instabilities which are likely to develop during a very small 3

break I,0CA. The bubbling of saturated steam thorngh ,

+

1

~

subcooled liquid and the injection of cold makeup water into s,

.q a steam filled cold leg pipe are inherently une. table ,

.J processes of particular concern that need further -

g consideration. .

s. ..

1 -

1 .---

py (... ,

,g - - q ;, . ,

. . . .u . ,

1 .

.u.

I

~ s .

(- --

3"g'.-

^

-- : - :; O.,rJi,

-- . . .. fr

((31.:r. ..3 3,. ,,

l O12G "U"

e

_ _ _ . _ _ _ _ _ . .-_.-.-e__. - .._...m . . . . . . ,_,.._,.,,y _ - _ . . _ _ . . _ _ , , _ - . . _ _ . _ , , _

g o t.. . r F.. . a.

.e. -

best copy .J ', -

.,*,...-._.w

~ ^ ^. . *

, m

'; e s e. *

. . . .30 .,

..+ ,.,.

0,;34 '

..e ** O _=.% .,. f* . ; N .

.g.,,*,./..,

./ TABLE 1 -

s.l *

< ,' 3 .  :~.?.

x.

.e FIAW ARIA ColtAESPottDING I.'.'

.'.'.a

- . TO ICMINAL PIPE SIEE

• • ,,y 1'.

o e

loominal Pire 11Te f f e.t. Behedule No. Flow Aron f *t ti e

~. . . a .

. 2 1 00 .

.0050 *

'1 l . -

.0098

-- 1 1/2

• 160 -

. c. y .~.. .x..:: ..a.. .. . ;%;.% .,4... ..

/.% 2 . . .: 160 .0156 ,

.j' .

's 1/2 160 .0245 -

'f, 3 .

160 .0375 .

-< 4 160 .9644 .

v .

.i a  :

( .]

l *Pctential single-ended circumferential break area. -

? .

U xg

. v

.2

, . . . a. ~. . .

- . .: .: . t. > . .. . .f. - .

t .

• i j * *

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

a. . . s

.s . . .

..r.; ,-

d . .. . .. .

..

• f." s f . : . .:.2.,. .%...~;!'b.*t.. ., . . . .ic'. . ., :. 2; . ,

.94

. ..p . ..

4

~

. [!3l,5..

- , . ,5..',O/ F.- : .

l 0126 ". ' 'T*'- E .'

- --- .:-- w. -

e 1

.-,,.4

q.t,,,r.,,:...,

DEIO - f,.'O'.I' !@; ,l :

e

--- - ~ ~ - -

. . . . _ , ... . _.-= _ c __ .: w - w a - . -- .t . -- .

i i ..

....= .

i .. -

.. /. [  ;

l .

be t.

1 1 g

...g.-

4 ,

• h II J -

i

. I. _ _ _ _ _ _m lt rJ ' Ii 3 r'  !

2 i i s

- 1. f.

i. l. ". -' . ,

' l lo-l lbo e- Q,, sli. : . , j ..* --

l.

~- v y

.. .l

}%l 1 L y ,I,. i i d 4 31 l .q *, _ .; l' ..

i

.or 1. --  ! ,

i e 9 l

--.._4 1

l p. -

( --

Ibillif

.i+. .,,

i

_ g gt 3 p' "'"'

i  ; . I L -

I 2,,", _

. .'Iy n

%n- -

F= -

, w."

., i [ -

.4 y .<

, . . v. .

. .. g .

c.

. . .. a 1 ;_'- . . .

..:. . .:g;.. .. . -. . . + .; . . ~

t

t.

...d,- . ,. *

, f _. ya . . , - ~ * % '.

. .i Ettan 1enn k *~ '

J

,. --, - , _ _- , ,.~._,__, --

__,._._.,__ ____._.___._,.,_ ,.,_., ,,,,m,,.,,w..,,.,._., -

. . . . , _ _ , . . _ . _ _ _ _ . _ . _ . . , . . . , _ . _,,,m.,, , . - - . . . . _ - -