Text

Forwards Safety Assessment Repts for Listed SEP Topics Per 811103 Telcon
	ML20039A385
Person / Time
Site:	Oyster Creek
Issue date:	11/04/1981
From:	Nagai Y; GENERAL PUBLIC UTILITIES CORP.
To:	Fell R; NRC
References
	TASK-02-02.C, TASK-02-04.F, TASK-03-03.C, TASK-03-07.D, TASK-05-05, TASK-05-12.A, TASK-15-01, TASK-15-02, TASK-15-03, TASK-15-04, TASK-15-09, TASK-15-1, TASK-15-11, TASK-15-13, TASK-15-14, TASK-15-15, TASK-15-16, TASK-15-19, TASK-15-2, TASK-15-3, TASK-15-4, TASK-15-9, TASK-2-2.C, TASK-2-4.F, TASK-3-3.C, TASK-3-7.D, TASK-5-12.A, TASK-5-5, TASK-RR NUDOCS 8112170473
	Download: ML20039A385 (152)
	v • d • e

{{#Wiki_filter:- >.~ t PoR.Q. spo A158L AWIS df GPU Service Corporation G s eWEe '"t*'o*c' " *

Pars:ppany. New Jersey 07054 201263-e500 TELEX 136-482 Wrter s Direct Dial Nurncer November 4, 1981

? to g b' f * ~ Mr. Robert Fell U.S. Nuclear Regulatory Commission ' -d DEC4 1981> 8 79C2 Norfolk Avenue L' m e-uxosa a g,o*wmW / Bethesda, Maryland 20014 s M

Dear Mr. Fell:

4 'LT As per our phone conversation on November 3,1931, I am transmitting the enclosed Safety Assessment Reports (SAR's) for the SEP topics listed below. As I indicated, the SAR's are transmitted directly to you on an informal basis since they will be used as input to the draft Safety Evaluation Report to be reviewed by JCP6L. SEP Topic No. Title V-12 Water Purity of Boiling Water Reactor Primary Coolant II-4F Settlement of Foundations and Buried Equipment III-3C Inservice Inspection of Water Control Structures III-7D Containment Structural Integrity Tests II-2C Atmospheric Transport and Diffusion Characteristics and Accident Analyses V-5 Reactor Coolant Pressure Boundary XV-16 Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment l l 8112170473 811104 PDR ADOCK 05000219 P PDR GPU Serece Ccrocrat:en :s a sacscary of Cerera; Putre Lia.es 2crocraten

4 SEP Topic No. Title XV-1, XV-2, XV-3 Design Basis Event Evaluation XV-4, XV-9, XV-13 XV-11, XV-14, XV-15 XV-16, XV-19 If you have any com.ents or questions on these SAR's, please contact me. Very truly yours, -- [ b / Yoshito Nagai Senior Licensing Engineer 1r cc: J. R. Thorpe J. Knubel I

i k * :b o Oyster Creek P Topic II' 2.C Atmospheric Transport and Dif fusion Characteristics and Accident Analysis INTRODUCTION The. objective of this section is to determine appropriate dispersion factors (X/Q) for both ground and elevated level releases from the Oyster Creek Plant in accordance with more recent NRC meteorological dif fusion methodology. These dispersion coef ficients are used in the assessment of offsite exposures f rom hypothetical accidents. METkODOLOGY ~ Table 1 and 2' summarise the dispersion analysis performed for Oyster Creek using the direction-dependent methodology specified in Regulatory Guide 1.145 for elevated and ground level releases, respectively. Calculations of X/Q values are based on probabilistic techniques which are one year of hourly meteorological data collected on the site tower. Values of X/Q are calculated using direction-independent methodology at a 5.0 probability and at a 0.5% probability for the direction-dependent model. For each model, separate cal-culations are made for the Exclusion Area Boundary (EAB) of 1358 f t. (414 m) radius and for the Low Population Zone (LPZ) assumed to be 0.75 mif4s (1208 m). Additional assumptions used in this analysis included. Lateral plume meander (as defined in Regulatory Guide 1.145) a. b. Atmospheric dispersion conditions are computed at the EA3 or LPZ distance in the direction of the wind using hourly meteorological data in the WINDOW program (Reference 1); c. For elevated releases, the stack is ll2m high with no plume rise. No fumigation has been assumed to occur. The value of the peak X/Q is conservatively used at the EA3 or LPZ regardless of its distance beyond the EA3 or LPZ as appropriate. 64 ,e w eegungemw m,hegemumpane. i J

f. -

Topic II-2.C Page 2 d. A building wake factor has been applied (cA = 400 m ) equivalent to i the smallest frontal area of the turbine building; A one-year period of Oyster Creek meteorological data-(7/76-6/77) e. was used for this analysis; ( f. The averaging time periods for this assessment are: 1, 8, 16, 72 and 624 hours for the EA3 and-LPZ. RESULTS l The results of this analysis are plotted.on Figures 1-8 and summarized in Table 1 and 2. The 0.5% probable X/Q for a ground level release, 7.7E-4 3 i sec/m occurred in tne southeast sector at the 414 meter EAB. For elevated 3 releases, the 0.5% probable X/Q was 1.7E-6 sec/m in the west sector. The values of X/Q for the LPZ distance are also included in Table 1 and 2, and are computed ~at probability levels of 0.5% and 50%. The 0.5% probable X/Q values at the LPZ (1208m) are shown in the summary that follows. In summary, the fcilowing dispersion coef ficients are appropriate for use in offsite dose assessments for hypothetical accidents: 4 ELEVATED GROUND TIME PERIODS _ LOCATION X/Q (sec/m ) X/Q (sec/m ) 0-2 hr EAB 1.7E-6 7.7E-4 i 0-8 hr LPZ 9.lE-7 9.6E-5 l 8-24 hr LPZ 2.5E-7 3.8E-5 1-4 days LPZ l.7E-7 2.lE-5 4-30 days LPZ 2.5E-8 1.0E-5 REFERENCES 1.

Woodard, K., 1975:

" Accounting for Wind Meander and Site Shape in Probabilistic Atmospheric Dispersion Models", Transactions of the American { Nuclear Society 1975 Winter Meeting; ANS 22, 365. 2. Regulatory Guide 1.145, (For Comment), " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants", 1 August 1979. -n

o Oyster Creek. FIGURE 1 Dirsction Dependent. 10 2 EAB "- Ground R.elease.s n r. . - u. n n l. 3 . I,. .e -l . l .l. I t t -{ - l I, - t . - =

0 -*

L . j: t i j l- -a-t .). - .; t a .g.

p s.

.l... l

.a.i.d. i. _ ~.:t :. l -_- :l:

'l. e.! l l- ..g . l. c: :.. j_; _". l. ;.. 3-t... ...l. _.:... ...s.- - -. l._... L... i..... 4. .g. i, i 4 f T. 9 8 1 ._.,. l. s_ .g -. ,.. _ _. -.e._... t 5 I l~-._I.-.. y.; - e d.~'. t ..p... 0*.. 2 %... ... *}I 4

e..-

i p ,..t ,..l..'. -3......... 6 ,.= .h c y.

-. s6 t

p ,.t 4 -t'

I 4

l:..

i..

a. --_..:r.-::.:: . i. :. .j.u. .. q- - . -.. I : -. =..I-i.- i i . ? *.:.. r

...s-

..g.. . :.l c.s u. - - hc......: s r:---::. .::: - g ::

u : :-

-- e :: -

...t......t .......1...............!................-. _ s ............T--.. ._...L. .............. t. ...)-.. ...1..*.... ...,.. t.. !. .e... .....I..-. l ,.,..s... ...t............. ......l........... ..s .....t............ ..l.

,4....1..

....... k. r 1_. -t.~..a.. 6.. ....l.. a. 4..-.a. ,.-....--+4 ...- --p... .*~l ..{.. i., ...;. }....}. e 4 ...t. - - i .sl. 4;.__-.t._ ..:3. s...y _. -._._4..._ i- .. t... .... ~.... < ~ L..i. ,... b ,T---.. ..l l I1Our /.

...q. -

._7, ..i. t,..- ei.. ,,.y-..._ ,g..i p 3 m l i.s,, t, q3,. 10.3 .g "j . :N .:.f., !.: 4 i s.s r.

1..

.1 s,. ', t -. ~ 4 - 4.: - y. s t : ---- t, cg it '. t.s i 2 .a.

.~.:.:a e..y.._.:. :

,t ir... e i n.. : q.

i... - i. :

.: I. t...i n - -- -i : p .. _.p. r.i s ,:....y. _.. ;... ._1:q. i.- !..i.-EiF:i. i 5--- E+-i- ':M. y 2,. :r--h ' !. = 1" ..7-'@ 5.r. ::j..-ih .J.:.-=h y i ;.:}.f_ f:(. ._........u..r.--..._c..c............:r-.........._....2...:.:._;: 8 hou m :.r. N:....._-..__-..:.n..z.:..

c.n__.c

....: :. :... :..... t...u.. :. t 2..._. .t.... n t.-.,., .s..i _....._M.s,.. .. pr._ _._4 r .. -. c..,i...z....... - Y. - I _. q...__4 ..,,. 3 L._ I. u-:. .s .t r .s. a. V es u.. -e ..u. -t. n n - f t s r + .n r : .t. s.. $._... <. s : 4 . l. ..i7..1..)U.:. C.: :... 4 4 - ___h. n. _. _.. ; (. - ' ;,.i. -)

s. ii Lr i

. w 5i: m '. .: :.m..

r. ' * *

- 3:. ... a. _ :.. p.

-
r.-t W P.-.

'.g~ ~ M - C *. L.' 16 ho u r s. :..w ::.. y r.2--W. ' 1 -- 1 ~;u. t..L.W - V-~4 ;- ; I -. ' ' M.-: _'. X n ? t '. - -..._. c..: :: u..._ : 1.. n..... s ::r.*..u. 3: 3.u.~ H.. :..v, &_. :.. =. ::.:.: =.. :... .?:... ~~.. __ _..- ~..d. 1 .:-:.~.........--.-.--.. .u.. ~. :r.. ; u_ _.........a. ;:: g :. i.._. ;

.m..t.=. _... =. _. : p : w r:

..q._ :. :r.. : =.. :

i. y :tt

.... n....-.m..._..--7:!-

. ::r.;1.__.

._..:..: _: :.:.. g y. ...._._._....t_.i._............_._..._.___...._.._.t._...._..._.._._..:_1.._ ....i_.....___...__.t,__r._.._.. _.r__._.._.._...._...__.... .._......._.~._..t....._.. n .. _. _. _ _ _..._._..L..._-.___ _. s ._._.z.. .-i.. g ......_..._. _ r..._.. _. _._. _ _... _. _._.._ _ ...L,_.._..___...._..._..... _. u, .___.._._...4_w.__..L_._-_ .a._ __.. m 2__ t_s...... \\ ____._._ __........._..j...... m.. _ ...__.4...._....._,_._......_. .... u.. _..,L_._.._-. -... _.:... 1 __ w.

.. q__.._..

__.i _.._. __ i _ p. _.1 .. a .a, ... _ _ q. :.. ; ._.. p. .,,_ _.... 4..._.. j _ j..n x w g _ __....g_.a_ _ l, _ p.__ __. _ z.. gt..., _ a. aL.,... 1 O.4 -_ e -. _. se. .i.._ . _. _..i. _..,_.i %. i. 72 hours - s , i. '... t ...p-Di a. - s + n ~\\ c m.- em. * +- u m.m-s, e v._.y.. y... m. -.. N i..,.... s. m...r L..s.,... + i _..,., _ _....;... -, ;. :.,. i.. m....._. ,,, I:"._W.I m.v.- p..ir..- N.i-i M *. N. _! .. r }." i j f.-

t- +

c:; r ::t. Q.:2..jF.-7 : i .p.:.:3 p'-h. A d3.'. s_ .... _.. _....._.._r._..._..:;;:l... .r ....s...~_.:.:._ .. :. V......:. :....._ 3,:._.......__.. _.._... --.....;.:n. .i. ....i...............i._.. .. }. ..i.:.... ....t.... 4. ._.g... .... -i. 1. i.. 1,.:., L.... -.... T ~ --.,~ .1.5_.,~.. ..i,,_.._.._.. A..

t..< -.

. s.. _x ...- ;I:.

1.

m. 1 *. .t . I i ..i:.~ a 1. - }.. ..i n.:t.. . L.. _ t,.. ;. **

p g...J -- _.. :. a.. :.

g.. e 4

w. y. _. y4_.. _. 1. _w;

.d. . r. ..a,. m.2. n. 4_ G24 hours .- w. _.=. %_;- A..v.- a. a.. y :.... L~.. .... 2,_;_ __ y. :. ;.._..,,..,a..,...;.... .. ~ -. - - - 3 u. = m

c c..:. :-_i. a:. :.;: :: I

_.. I.. l......-. p :.... :::---:.ur. :.-- h.g-t,A,1..r. t - ... l - - - n- ; f_y.. t. 8. -~:.m- -r- - t - . ; -- ?.-

u.. : : : : ;.

cv-- --. ; 1 _.._......_....,......._._._..._..i........_a.._........ _.... _ _........:.............._t,.. ...t._.......... ___ t. ..t,... l.. ........t.. w ............._.___.....i..

_.,.1....._.__....._..i...___

...,..,.._.[. .......!..._..._.L...... _. _._. .t....... ..g. __L_....,_.. .....L_._.t_...}.... . _ _ _......._.i .t__.._. t._ .4.....,{._.___.......;.....,...._.......p___1...._ _..t,2 _ m7 -. _ _ } _. .... _ l = ...a.._. r _ _ ;. _.,...t.. _. -[-_ _._ _ __;..;..._.. ... l. _ ._2_ g.. g. _4.._i. ; -..j.. .I . _. _ _ L., 10-5 ..._ p-.,. _. t....i_ _._.m..._.... r - a t,, r, ! -.. .st._ _ " h.....-im p-- m" - -;rv r,,- i r q rr-r -t an .a.. u u i i a u - o >. a n u a n u u

n....

Probability 3

E t 0 star Creek Y Dirtetion Independent FIGURE ' EAB - Ground Releases 10.2 w..,. n .6,'.,, -i....i i!. .i

i

al 1

i. p i.a.,

i j ":- -l t. t ..): s.. p,. l' .v. ). l? 1 dt. 6 e-l. -j '.o-.[~ p .l.'-i..t- .......l. 1-s-.. . i..: ! r p .. : : =.... .--.p_._.... r:...: ;. . n... i.. u. s l,.. .t,-...- 1 t.. e --, -;. i r. . i. n_ a j S. . : I-

[

.l) I- .b.t .4 . ----i i ~ q' -..[.'.. ' ! ' .E-l ?, ] _... I.: ,s l t

a. ;. 7..-

4. .j; t-t l,, l..{'..,j- .a. )

6 I i

'...?. 2 . ^. ...."-'".I.. l.~..."***U. .. " t,. .t..- i t:.: - * :*

4..

i '.

.........\\... ..

l '.. : '..

4.. -.: -

....j ; *.. 1 .l' ..8r*. : :. .;. ;- ! i ;p { ,. r. t.t .x g.. .P.......... "..**- "!"""...~.".T.' .g e.... . 1.-. .-...,.t... .1.. i"...... - ..f 3.-.. E** ...<......-i...~......' l --......o ...._~..I..--..I~...- ...I.. .l..!.=,.... 1., -..I ...I... ."I. .s.... ...... 'I... .....i. ..,.. e.... ~* =.... =.. t ...,.-..o..-...- ... +.-..I.....---9.-.-.- .. =. =.. l... -, ~... .g L- -... .. -..l... .g .. g i.. 1 hour 8- - ..........a. .....--d..- ...-{-. ... -...w. .{- -.. -. . - ).- ..l ..- -..... v.e. - r.-i.-~.a - -- -. g,.,. ...l..,,, ...I ,..I.-... - L'- --p .I, ~ .......... _v.. t.. ..i, - ~ 1.- - -. L. - .e. 1 g ...=i. .,.. -: :, 3.. i. n i s, ' r t, i ir r-7 m.;, 10..3-- m

l. m. a.nr.-s-P

+m. i r-- : - m L -- - m n-c- -r-e., m..s - '. R - m*, w .unu:m M. -. m. r....i. m....v-w e. .I. + 4 w c m._:. 8 hours _.._.. g..I,y. m ;. j. z _ a g. g. g, _. m..2_. u_.. g :.y :g.agp ,~ -n . t..... _.._ :....._ m..._.= :._.::: _.. . n. :. : ; :._.... :n... e... _=.. w: _.......l.......... -..::.n...=.. _=.._.: :. -. : :.. i..

x. _.:

4. _....,.v..f _. :,. 4- --m..r. ............_._.t... ... M. j. .p...l y.. t* l_-w.g -[s

g..... r. l. a.,..,.._b....a_*._s...... L p.r.. _

g . _1. . c - i..,-.=.- t 6.%_..N t. s. _.,.,N. L . e..u.. - 4 . -e. o c a r.e .g . _ s. i.. q . p. u. {r. 3 r e.. . -aL- ...e. i _.. t 1...Q. e.. i.w.J.: .?,. a- -.a.....: 1.. ..t L 4 a. )..? '.d. i r .y..

.,._......e:

'-:b. m ?. V.W.u-.--i. h.c:5.- i G .-.m.w- - . :F r.-n p -.i: 16 hours : - k D,: c4 -r-r. .r t.. - MsEdd--.... -- J.:.& ::sP :W 2 8:-.:.I. ..v-M - =.i._..= 3 !F.- - C ". 6 4.: N.a :Ai.f.:CT_'.' l

,..m, - :9h* L

^ ...,... c.- _-..:...-.. t. r.-..m.[ .___.1.. j... ;.. &....._.....g. -. _.. _. _...... _...... -r... n_ ,_.. _.. t. :. s.2. _ M _=_ :: s.c.r....,...u.-e... .....g..._._.,.... _. _. :.........}... ..r_--_x.-._-...-1....---*.-..3..*.__ g - ~. ..g... ............s......t_._.,.._-,._..t.-.._....1_..___......J._.... _.. _. _ y _..t...... ..4,__....., _. _.... ....t...._...,.... ....,.._..._t._4._..._-,...... _...... o

s. _..., _.

.t cn ..............r_ .......,.......,...____.._..............._.C_ .A......___,_... .-.1....l._...... v .. _,......6-..__t,........_ ..m cy 3,,,, _. 4... .. A _- . 3..m.,4. s... d......._.3.. .4-g ...e. J _.. - %_ .y ._~_,%.,_.... _ ! 4. 7. ..t ..p. .. Mt-- -. r --- +-. - l....-. a- --~4- ! -t ' '+- - - 72 hours --. N.--i-- , -F w.. t'..., Q.7 _L._ .l a, M. u N.,,... L. .l. .s.... ~:. f- .... ~. i.. i.. .N..... ..,.__i; .i. _.. h.. %.,g.... .i m. . _ { _........ .. g z..... ... l l.. + -4 p, ..7 s. L ..--P -* - '. J : - * -- - 4w# r ~. 9. - 34 _'-q . m r..; - 't.m- . h. : 6 .r. . bet,. -L.- m: ' l ' i h - - p :: r. .l :..:*-t i.. - j. 2 W th, i.: : . u.': s. -rs l-

."t l N

.t W.1 I o ..-. s - - - - - - .1 E:d.. --I.':: dM: / L t.. s ~ W h ' 1 T- ' M-d-M: 8.- 1 ;\\ W ip-dim 2?j li j

~ ?.t P '

t..

w. :. a s _. i..: . 2.. :._. =_ ::_. ....,n a... {;.i.,..:. w. s........ N. : ..........,..... e.:. . :.c.1.....,.... ...... n. :. 3...-.. ....[. t. 4,...... i. .t. -6 i -- 6 --........ t .., +..... . i.... - t-t..... 1, - - i-- - .i r : - A. * - ") . ]. b....g..---s.-- .g.. - _.. ;,. 4.. - I.. L.. ---..u. m 5. i. t.... , - -.. v.a :. .. 3

h...,. _. d _..1.. a._.

s.. N 4t. 1. ,.... _.. d...- i

n 8

.J. y ;.m t -._;. :- :- h4. . !--4. %. l.. *-. b.. i ' I 624 hours . l.. .. : [.N l. '.a t.h :.wl. - [. .- i ::. e. t.

L.::.u'

.. * - r - !... -t----- .~: ' 1 .sw.:..' .. ' :. :. a ..,....._._.I.4.s_- .[......;.:

....~:

-( :~ ... i r :... e - l,

i...

....t .1 8. ......1. 1 ,.4. ......._1.,............. _~T._... ..~._......l...._.!..... .. l,,,..... j..... : .......l. _a.._.6... l... _....... _. _. ....{ ... _.,.g........ t. ........_g. . ].... j....... L... I.. .&=-.-4+-+.*4=** ,.............;....,..I......,.7,. ....t... .w... . p.. =.'._.g_==.9..g...=.=.f_.*-+-++*.--. i. 4****4+ 4.. ..: 9 4..

==.4.-.l..

+_--h+*-a-. .. g. ,...-.ta.... + g :.. .=i ...r,..,.r 9 -i_ .....,...._s_ _._. - .... f. f _..._ a.. ..,... h. _.,, r 44 go. . t., _J. .... -._1.c..- ,,....i_._ ....._..e.

i..

...r. .I...I.-_ j., ,..r ... j. ': i.. ....l.._. u.,,'. 10_ .n ..i n .. i a n .u o a u = n u u Probability

Oyster Creek. Dinction Deptndent FIGURE 3 + LPZ - Ground Releases p' "- g ."p 10 ".. __---t--

i. 6

,q....".'- II l l--l* b I l-I- I I -0 8-1: i q 4.,. g .}. . -i e. p a l y..l-

.p --g..

l... l-q .L . p p: q -l T L, ,, p .._....i .i.i.............. .., a, . i ...-.i....., t..... s_1 . r.... l. ...ii -. 1,.-... 1, 3 i ....I. 4 4 y_ .i. 4 . cr a... . -, 4.. p-. b

,;.._'._ p.... ; f.....

Ia- ,n .i i. e-e . m s.. - e l. -..g.. ! -- .--r.4 y7 . r-- ,: 4

u..

u.: - ! ...,e...,, : n.. u....: ;. v-q ? w !ie. i i.._- [ii.l. % 1 8i_- 9 ;. P._.'. i

* ~

T.~'m~'.~J E.U: p-::isiiQ H O. :l:% i --s ........-: :-. +

i..

.....----r--- .. l.............-..t.n.... . -.........c.....1..._.

1. hour_..._........_..._....e........... _... _.

_.. L, .- i.. ...+i. s-1..A...l-~:..___. ..._......l._..t...................._, ..l.._.. !.........._.l _._......:..... ...l... ......t_............. _. _... -..... ........... ~. _. ..._g ..............\\.t............._._...-...........-..j....-.l.. I. e~-.~+ ..~.g..........-,-......-_!........-_g.. -.- ..(.. . -.s. s-- - g. i. p. 3..... -.....,........A.-- . + m--. 9..--,.---.- 4 ..._e u, .t -l - J. .w ...a....2-r--- +! 8 hours .f.. _;.,r. f. i r.....

t...

4 __._. 1 ........ i. .V..x ..., a, ,- ii l.~,.,. p . _. ~...... -4 -d.. .,...-- -ni y ,,t tm l ' '- 1 i ~, d. i ."1TT 2 .10. -t--, i j-- ;...... c... ;_ i. c. _. i. u... -... - y., t r.q., -. a.N. u .:. i .g. 2 g. u.- i % >. .. c. e-. 1 6 h o u r S :.=.S _..t._. g.-. -.= ;- - J a.-- -p .n m._. r.. ;- u..i.-_.:-- - u ;.; :. n -,- p-r. - . _ q.v..y _.;:y;.4 yp t :i-esig. -b.m t i.u.+.m i n. - :- : ::t :2-n.=.=: = :. i :.E:n' ir:-i.:i; s_q. e._ -.- : - jq : a'._v: *.._.:.:_:- _=.: ..n..::...=.. i...w. :. : = 3. :. :. : :::.: a = ;:.: : : : =: : ;.: = :. =. =4 :.: :.- =.i k..:..f.=. y-_ ~~ =..s.. .= :... s :. ; u.:: n-.. s- .l. u_- r - -- ::.. _. - - ; n : : -- - u x :--. : -. :-.t.- =. m t l ... rs..s.9...6.-) ~?. \\ u.;... t... ; t..g.s... -} _.+-: g w e.s._l. 2., d...p..l., ,--:,....g.~ r..~ sg.,-t.

_: .. q..63 :g... ulut g:.

sc .2 - 4 -. 3.. 4,.. ~ M :... - .s .,z.n s: i... s 1. - : a .. \\..;. 3 ; s.....: t - 2". }.. q.. - ...a}.;..i....s..: .:..... _..s 4 j.. f d s_.- . _: - i z. g r.

..:. f....

-. 9 $ l*.f [..I *.; Q * ;*

f..** 4.( I C.. ? m~. ) [*i.. ~. F;-*} d I.;* -
h h,8 I.N h* '..".

m

r I I ' **. v

~ .:: r. a -. u 3._ = h ' }f h.*. -l.*l- ....I.. *\\'W *'"* =.. ' _. - -,, E:. 1 F .%? g w:: ._n.:_a _ 4. .}._--b.:-.. ..a c.. : Lm ::.*.: . :-a :-.:. ;.:.. :.-l .. -..y [..: :m. e s q : g_. _-- : - -- p. -: L., m.:._.r. r:.: ::......,._.. u- _i._.- 3; - a. - .:c . _...._... :: ::.... r: =. : : _ p=. = :7. :.::., _.,....=.. :: :.._-- n =:_. ::r -.. r.:. :. ._=_G' :. :.v._.::.- - u ~ ..}. -.....o~.~n-- .. ~ -.... n m 72 Murs ..,...._.__.-._.'..._....t._...-...._...-...._-..._....._-.-_.--.._.___.t..

m.

....,__._._.s.. .........?-'A.,_._...r. .......t.... .._.. r_......._a__.._-._ .... _............. _....._.._...t t. _ .-.... ~, _ _... cy .__._.g._.).........._..._. .,.__m..._i.... g ._. ) x . _......_.\\._,.. .......*~t_....._.1._.~.,..,._m... y.._.__ ....p.. .[ ..'t__. ( I_ i.

q....,K. ~

.5.....g...j.............. _,..... .y .__ _.w._ ,. 4 .p- .. _. ) _ ., p. ..g ..... _.. 9 ~..I - _ f~. q.. _q4.._,,- g.. __..}g4 ,. m _. s __ }...__t__.._._.. _..t t.'i. .N_ ... _. l __.__.a . h._. L. < 9 %.? -. i. .\\.. .:.p. i . _1 ... -...,. =.. _ ,..... ~. .. T.,. _.. 1. _ - .,t....- i..,. _.__i 5 e . i. 10 .......l. l _...6%;_ _. .%,s i n. .i l ) L m. . 4 ,\\ ~ )i g _ I 4 l. - l.w s. - f f.%,~. . s s: g

t,.- m.~*.6

.u;.s._s?t.. s - :.e:. _1 i.n a ..n...- 3. a:. li ...:2.- n-. d a. ::: -... :.--%. - r.;_; k 21:. A.-~~-T g.,.:_- - I. :... '.~e.: *(. ' ; r M. -ti .t. I 6 2 4 h o u r s. F i=_ ::.t t. A __ n._ A ! ..b i..v =- ir. t u._ : Fus:+ +s.T '.V M.-!:- tf.=1 - h 1 _v. _

n

-n

.n r.:.:=. _ _ :... _: _ : ;=. : :. ;...:..

... g..... .... :. i. - -. :- :.:. ................. - - r -- - - - ' - - - - - N : --- I.- ' L *.. ' _ a, s. - y 2.. s.. .(..- p ] i t. i l. g . _. -

_... ? g _..

. Le. n. s

r...._.

t ..t i.: 4- -{.- e.-

1:.l

.L r a g-j..p=. -. .. -... - l u r..._.. :._.s...- g.~d. J...t ... L_

.. 4 :.._. -.., i p : - - :

u. p.. j. t. i,._. g. 2 ..'..a._ _....,.g.,t..-

7 r

.p-t. g. y. ; e..n ,n_

p

- -{.-a L._ - --.: _ : :;.;....g. . 1 --.- -8 1--

p

-- - kk:_. m ., ;M.. r -;.a.:fe._ .y.. . : rt_ r b..L.: -.-: _T.. i..... 5 2.. H :i. l..:.:.i. -iG.. : r T--E p.r.... .......t........ _. _. ,i. ............t..L...... .i. i .... !.. t. ....t..-.....1.......+ ...L._1.......,_.............,. _. _ _ _ _ _....... _.._L....._ ...... 4 ~.. __,..__1.!__.,._...__ 4. .,_ p.. -.._i. ..p...... .. ;._. ___ p.~.. _... {._ _.. _ _ _ _... _. p. -._~.. _ i. -. p. ..i..._ g. ._ L. _ L.,.._.e 9. _... u. _. s ._2_ .t _e

g..

_6 .. t ..t,. L. a.,- ..i_. ......_., _..i,_. 10 .... r ., i t, . i , : i........ i 1 un ou es es u s a n. n n.. u a n u n n n n s n> s Probability .. - - -. -- - ~ - - -.., - - - - -

Oyster Creek Dirsction IdenPendent FIGURE 4 ~ i LPZ - Ground-Releases 1 0 ',3 m. .... n u * ., n. !.: a i: le.I . r 0 l ~ : 2 .i t: .p.,. t l~ls 1 ..l-. en.i ... t. p i,. i i -.. d., i. .l.... - t, 9.. t. 1 -.. = . a----, g.:. j _6, I.; u. q. g..... ) ..r..---__.4 .l..- 8- .1 - t 7l :,,. + ., - :I :.! :.. - -* - mi..-Is. -d.......I

b: :

.) t~- ..S.. ! :.a:i:..,:.l :: . n. :a 1 r:::. : _u -i: :t ~_ l M.. - u s... - ..s.; u.y. 4.*.:.: r t:: J. t mi_.4 t-. S- -..-- - .. l -.. I,. _ I.- o ' I. .q 1 hour L. v-3...... .l . 2 9.. .e. i. j .e . ~s--...- g+ - .t.r..j .. l. :-.-. .M; t .r -. 4.. _p., is. au i. (_. e - ... l :.... i...:. l..@. ~--1... i. l :. i.. :. .. : i. '. :, l r ; .i..g. 4,.... ?t- .. 3.._ t.. [. i .- ;. t -.. -~m... ...__-r-... ..._ : e.. r.: ....r.

n,.

11---.t..- . -. 2 ...i...,......:: ..r.... - ~.: u r r-c:.: ...l... .......... =....... r .a 4. 8. g

- .C......*-..b.--_*

.. -. ~ ~... _. . J... ~.. *.. 3..=8... .= ....t...... .......1. ..l. = =. -. =.... ..8-j -...e

==.*.*.~.~a.,....- . ~... -.. ..,!*-.a....-.i...- -,.--.e-b"-...-~~9 .e....e-.e... .... t- ~.-- +..... ---.**~to"t.'. ~.. ~. - ...=+--.a.*.I'.*f.*-- -... ~. .s ......c..... \\;f... a e L--=. ..-4= . t.. e a .....t... .... = ~. ....s.-...~ . g _...... p.. 6 ......._l....... -.f +r F_- .-*---I-"_ ,-d ~" _..-t_._-_9 " 8 hours .....- p-.: s -. . u.... .a.. ....... l... ai L,, .... l.... ... p- -.n i....,.....,...._..,..........r..~...~_..,, i' i .e., ~~ .y.y.,m... p-g-g T. . _.. i .4.., .........r ). ... i-. 4 ..t.. r . ty.. 4 10 t__ . :. G :..i t .s-:. ...a ..a..n-v v J. s. w -s v.. s...:. _w : w.,.:1. y.w p..g..p(-- p.- yL.. :

i;. a..:. g..p..r.n....g...

...... p ; i.. w.:; a p[- 16 hours --t....._ :. W-M W !9-W h-r=Fs dN;W m*%:# "-t v MM - i h r, .. a... :

l.... )..;._ t,,. ;,. ; \\. t.. ;.,., ~; ;,.,.y,,;;,,,, g,, 7., _~;.,. _.,,................._..._...I,.y,3_.,.,;.,.,...

.. s,....... ...s_... ._....L..i... .1_.... t. ss.- t.g 4. .; +.. ':.s.. Q.... A, L. } - l :+T..,.. i e. i.. i4. t '4 3 .. f. 4 - rs l 4 g ......y..- _e. u : t,. !.,.- i i 4 l f y ;4:.. t.. -.'. _W._- s-q.._. -t* _I s~ r.I t(. 13 - ~ ~:. ~n s r,1 1 1 ~~. ~ 3- .. y-.. ,s;.',;. ; g - U. :....... ? _ t_ n 1 ~ * ,,.r. e h. ; p.- i < ~_ h 1- ...,.,:'i -".. K *. .2. Q:. _M !. M y,_. :.N.:.j(q.!.2 - y - 1im* * ~ * - a i-19*.8 s . -4 i n_ t ;.. .t..,.,4

a. a a -**

i .c. .a 2:. %.m2.; ;. a -.42.

1. i.H :::.r : -

,3 - i..i.= m. T - n. i. L i ?. ) b ** .,,.. p_: m,_..a :.. N**.' t. :. u. n_... lm. 6."_:.u.. ::_: . :. d."_:@...

g..

, _.. :_ _.w:. L - -.......m....-.t.r-...- ,_. -. p r-a.. .. t... .q. __q:~.:J =u =~1.:m.:.n...-:n :.:.+. _. -- r

- 3. x

.: 3...:s:- :..:

" q: bu :. :. I :....:.: :... ;, - ;..:.:=- s;:.: 5--a_..m:- : t__ _:- :

: r - :. - -=

....i.._.._.<...._.i....... _ _... .-_......._t...._._..- .L._.. o u.,. s .._........_-_......p, ......i...., ........... _,rt __.... r...

u.....

1._.....,. _..... .:m t.. .... -......._..L._. .4 _,-....._.m, a s. u. ....... !... z.... _........ ...._..m._...... . r.., __ - U L o w...... y.. [..,..,.. u.... _ p...... _e.......... ,..._ _., _ L.__. e .. +.......... ...t ..s...s ;. 4.a..m.... _.{- ..... 7 -4

p.._.

v + J.. 4.._g. }...

7. ;

7.__. r --- - - - -s. .o,, .... p., .4.. i _ _ _,. ...t. 6 7. x t,_... t w s j ~... e. -., *^ 1 ). .i._..,..... l_ _. =-. \\~..- .>.k

i L. i...r.

-5 ,.[-. 10 L l. _.%- I.'N_@. 4 t_ - F._ i g . ', q..u i i .r .. [.., r.. _ _-. pr. ;.. .r... .. u. t.u. t_ . p. a. 3,. 2.a _. .}l Q :..T:: .r.. ; 2;;.. l.

p.,.p (.. p.T.: ...y c -:j 2:r 4 i :.

~*- r-s. g24 hours ,.,.g:g. ; p. a. p_ i J

..gga

..: p.;.1 i re:... g a.q=::.g._ q.y..:. g:.,: _ i. ,t..

7.. _..

__ t. .l.:1 :~... i... L...,.._.::~...._. ...,_..j...J.... ...t... t. ..t.. ...1_...:_===__.;. :. : _. s... :..:.. u_. =. ; : :.~._:.

x..: .... n \\..... L.. :.:. =. : t._ ~ s...._:

r

w

.. $....._....t......_ ...~.. t. w g..,.%. t ..t.. g- ..,m._. .v.,. m.*

.s. -

__6 .t-. i;*

'.F y..s, j.

l-1. ,rt 4 3-jf... j*}.s ..pv..w..J;,....a... l _2_.nw... ._r a-. g.. :, . ; t.a .... m t r..... .q.. t... - a. s: t.. . t-v.

4..-

7;i.._ . g. m ;;~.y.-t P i : 't.. CI. 4 :e~M_ 4: _. _.,i...i r < <.. _-. i -w . :. t

t ".l.'. _. _.1 :.
g :*/

L.- :.; 1.:.?.: 3...t.;..* j.: ..-*.a.

4..

t 4.

.3. ........!.4

; I.. -.._.. _..

.-._. a '..a.? .=. t.: ..'. '.* *i...:.. l ::1. ; .... i. :. ~.. 9... !. ~. *. -.: :

4

'..:.*.*.....6.._"'"ll.__.f,.....{._~...r..,,.,r...* - *

a l ".!..* * '......

."i, g,... .e _.........._.j....._... ........___l........ ..... +........._...s.... lI _.........--- ...sa ..I... ..-..e.. .e 6- .\\. 6 .-ee .......I........-........-6._- -..... _.....)....,_.,.._ _...,.......,. _..4.. 6...... -a .j... .*I'*****~* . g...... ..6

* * * ' * - " * ~ * * * * * ' * *

...........c..L..,._"*'*'*- . _ '.... '... ' ' ". '. ' ' '. '. ' '. ' * * *I_

2. ***

...l.... _. a._.4.._.... 3.. _._. .. +, _ L...t._}._. t. ...o._ .,..__t_..,...._. i,.... .r.,..!. -....r 4_,..._,....... ..:.; } ..r.. . _.. l, _., r_.. s. _.s.._. _........ .._.\\ .t. u... a.. _:. _a..

t _u. y.i..a.. m..

.n ..E..._..__ .t L. _ ... c

a...,......a.;...

- L, i -6 i. ,.6 . i. t...- + _.,,. i. i.1,, ...... L ...,.. ~., 10 .r. n .c o.. i an .a n u u i a a u a = n n a n u n mm r Probability 9 -u m. .-p m-.. - i

Oyater Creek: Direction De endent FIGURE 5 EAB - Elevated Releases 9 "., j" ~~ 10,- - T-2 0 - j i , i i i, u.., i. ~* +- i. t t

1.. t i

n i i i b s .p. .:.t. .e.. l : .. l: ! :l .I - .:d

I !!

a.

i" r.

1 q--.r. -~. ! . i.. .--1-......i, n

s..

..--.s.- ... a. u ~ - - - - - -. i; I- . "T - 4. - t. +M i I i. 9.- - f, ! :.i _1 l, 4...

.i ;.

M.i.. e :iF 1 1..

- -l... e -~. _: r-

-t

g;;

- :I - our ~.- ... ~. ..........g...-..-~.~..- w. ..**-~ ~.. ".*.:..,........ .. _ ~. - .._..I.. .w ..e- . _ - _, _.....T,....,.L*....,..--~... - *.-.--.-.I- ~ ~ ~ ~ *

~~

w ~ - + = t . -...- ---~ ~ *. .. ~.. ".-....._..s...- t--- t --.t-.-- '.--.-e-b.---*' L -.-...,...t.---- ..r.*.-

1.p_

....[. k ,...1. 4 -.....-=--? ~- a t - h -

.a.

--.t..=. ~*t - 4_ ,..i- 'p... .4=..-, %.. ,.s,..

y..

,.e. -*~ e i... . t-.....-.. i I i. _*J -- -gn u-e 't* a.e. . L.l I ' m,-. w....,m :.

m... -

- r--- . m 8h -r t-

a

.. s .-a ~-- .. -..,e.... g. .s

1... e... ~.n.L*.

- - ~. ~. 1 0.6.- .v.~...* .e g

r..

t,... ' s l *. .i. 6* 6.- e 6 - L. ...1 -.. r.:.L. : -t., ear 7. [- t. t L. ~ t-.... .. ( ?- .- a .s L...:._-- ..t rv... .. y x. i. s _.....

a. _

a c. v. A w. s.g;.... p.v .7 m.=. .y..n p . -+.: i.;< ~ . p-.. A:: }.g -s .4 .iq!. % %q.;.M-i .Df. g.w .-E w.i- +u p _k m w. ...g* ......_.. 1.= : :_ _=_._.... :...

..a. = r :. : :=_

..._..=.=i.L.:._....... .. =.... t

: =........;....

1., _:.=.._=. :.. v u._..

4__. .._ u. :.. : ...r. _. 5 -z.. e s.* .e (. _...g . u.. .s.%.

2. s.. { s t r...: !.. u t,.....,I

.-.22p._....._.,.1~...3.... .me-- _ 3_- i. e_..,._..e. y_., _._.

i e

-r . r..u l._ t r. 4-S. _._-2._.-_ 4 a. -. ..s. .m: .:.- d'&: im..I..w... ..d___'..:.._...,'- ..w...:. L.nm. L..1_., :: t. .:.t i ,i.:i _..9 2 :. Ei 4-h.' '.. ' : 5 < 5. ~ s:..: - m .a aa :... : ~ ~ .y t : ' : 9 ; i - iM.'-5:: ..-,..;- -v:I *. : h '. i :

. j g,. j *g i, - a 1 y... d.

en ... - - * * " l.. 14.:_ :c_a. ! : L. ;-- :.?.: /. _....M.....,. : _.. g. .....;. : l. :...* ;;.n1:. : :. -14. : :

l:;.. * * *

U

...-.....? :.t. .\\ ..-*n. ;".* * ?. *. 5-Q Q *. 3 ..t. _..._......c_.._.._._.!......_.._.*.!'.-.t.T.t... .....1........i. .._...t._.......t..... _..... ...T.... ...c. ..t._.. m y i.._,____. __ v _s. _.._..t... ..L. ...g _ _ _........\\...h..... ....q.p..........,..... a.. -y...._.. ,.......5_ .w. .g.._. .%.....+.._. _... M

.=

4 __g...,.. .f.~_. _ ....u_.,_., ,.$... g ..... w ..p-. _.. ,-----l ...*.I,,. - * - ' - - '.*.b_.*-. ...3 w.g p........t.*.-'. >b. m m a _._ y .l 7 72 hours .. -.. N...,.. ~d.- r.~

m l -- t --** : T-*:

1- .i;-- m i :... r:- -

m.- - -t.. i.

. l....f . r ~.."r. p. m" .~,3 s. . i. , e 10, ~6. 4 g. , d.j. s.. e. M_ J N~..

4. w. p

_, m.

:. p. _.

.-l- . i. : _1

4. -

w -jy p, a ;..

p... t.

2.:..

F..

m. j. .i- - 1 a s.h M \\. f.- @ 5. - L , +.7+. -@. *: I c.n*..iNLS l :. !.. 1 " _ l -_ _:. ..c.. :. : r.... .... ;;;.;..3,. ..n.....-.. s. ......_i.:. i. _g_ ..s_.4 e ~ .p -l l.3.ia e_._. .. j... { q 4 _. -.l' _2 .2 l

.. ~ *

-~ s z ..i g i e t

y

,.i.

1---

..Jr..:... .: !...... a:... j.. .. ; ;..;_1 , - { a. ,g g;... , 3,... g.

a. J :._r: e....::.:.:. :..

,y ._.r._t..._.._ 1 - I - ...t., ...6 .__..l.. .s -....i. ......l. i. .J. ~ .t ..y_ ...-... _.6._._., ..l. ~ t.- - - !--*-4.. . ~.. 624 hours _. _ _.w .. : i... w. u ..t - . _ _..w... ...._t,... i.... ,. 4.... ....~ -8 ..i,. H, ..c. ..i. .i.. i.

3..

. en m. .. i i. a w a u = a n ni n. Probability i ) i ~y- ,-g gg -.Pw + - ' g-

Oyster Creek ~ FIGURE 6 .

I' Direction Independent EAB - Elevated Releases

'~ " " " ".'. "1 - i- - l - la 5 1 6,~,

,_,. _'"i__'":

'l 4 1- ', - Pa t. Lm. i6- ,. : l-. ". t. I - .:P

P e : - 6.:.i t i 1

.c ; - 6 e -t ...= i " (M :".. :- I,, s- ^ s.. =...; .b .ep. i u: S.. -..i.

........................g.. _.. _..,.....

L_ ......i. ---h-H -(, -b.-

e.. -

2 :._ m t- '8 d.M..i-. L - H -- {'i t - -- [ .. } _..q. -.-a....e - q. ~ - .a. r .i.. ..(.. . o: .*'N I hour 8--

a'

. \\:. : ala ..t' - - -l : :.,. ~; - .It : -' : :

s...

.;r- s ..:g-!di.d Dti,j.:H @- :j,:a-l?.:j . - H. I. *.. ; r~; e. -i: l.4 T31.m :l:':it;. ...t-- .s. ...?.:..-

fr

.: i.. ...... -.{. _ _.t... .-.-.,..._..-1....

n: - 6

[..-...--.

.. t.

.....-.-....t..- ~ ..:.:. t t. L. 3 :.... 1 3 l... -~.-.m.I............I.*.-,...-...-.. - S= g,

t. a...

I...... -~*.... .f.... I... _.... 1..., l, - ~. - ~. .-.t ~......t... .__........l ..e.... L. .1........... 7_0 v. .. L._.. ..I. o... ..- a.. lt - --.-n--- -."..-.6.-- ..I..-... .,.... l. ..........w }._.1-.-a!.~~..",..*- ...I....e.......-e.. ....l. _....]..,... l... .f...~- 4 .o.-. .4.. t-..*~ ~*t---*' ..-.6.. ... -F a t. - [. - - t 6 f'.... - S ..~t"*- .- E - t -I ?,. g gy*.g -l g t- - -t ei -.-.r- - - ~, -e-t - t -- _.,_l.... e - ---t--t -- - ..... ! _ -- : -- am..- --m r

t.. :..s-

w..,

. = _ ..........L.~.-..'-'. F.. --.l-. i ., w. .. i e. n.j .-*t,. .t u ,.,. t r .s . - --'T. , ( ;--.i, . d,.. . T*. *. .....q,.3 q y-' -F ;-- - ~ j i-- { - ..d. g . ~. -". %.- s., e a 6,- I >; .d _ e e. -..3 i m --*f-i. ~ ~tT ' P -l.. M..., .'10. 6 -,. ~ _ _ - ~ - i. -W;, ti.. ri w.* .u.4 A... i r:y p ip q.u h 6. ' ;....-4-?.i..j.. A. 8 : - -- W i. :1 t - ~ :.t..; *. t

..I....w. :t.-

.,...::"...ir. M. i 3-22. :.,., z: e s

d :. j

.[.0 n L :. .N'.e,. .. ) ;. ; :.7_r. ..i ; .r.t .Q.. ~ !...(......... "..... ".. ".. 1..r: : ci. r.:

l =....

>1r .- -. =- .. r ,. !:. n.... l.4..:l. ?..A d..r.. r i. _ _m:... r.. .t... ...._._.. t _:.. L...i. O. I..: :.... ;.. r :. :s....... t.

3.. ;...._ :..u :2. m _ : _ = =........n__

_ ; : t =:...'.._....~..f....,.1 y-' _...a...._....u :. '.. _ u..._ _.a. _ : : : i.. 5 m_ ..,.-.._...!c 4- ....., _. t y.. -.. .*.'T. .a ,.;w- .. F. e..9 e..' { i r.;..., ......]... +._ s _, : w i._s, N-- ....w. a,... . - t i - :. .. m L. :.. t,..... :..p. s=0,.. i... - d.,....s u :. . a- ,r e. e S- -i.m_._. r ;. :.w...i.p u ' : - [.. W. l..... 1.... :: ; -- N 3 ..:.i - v-i -... ..... '.0, [ _u. -w. 4pr .j g,. u.

-ij 5c.- i.in.g,1.j < -

4..):.. - h - j d. li.~. i F ' N : ::a 9-g G i, : t . i"-.- ui - Q Ud.+

..... 1

....-----r:----1.....: d#. . n. r.r. --- t, : e.. a.. c._ = t a.. ~. -. m...e..... :..._r.... L: : -.r.g

t. r. :-.- %A n:%. :-

. a.... $.- - l...... s. :..~.. x :. :l : : .r 16 h0urS E.. s. .. -...... !.D. I. :."~. ^ l.11... p......r._=. t.=. = =

u.. n...!..a.

. :. n.. u. _.:.0, E..__H. .'.. :i. L' l.' :_3. #' i i. . ? i_ i- ..t,__ . 6 .t_.. n m

s. _

. _......._...__-...-.....t-._. a, 5 ....._._.i...._.-.-..__.-_....._...._......_..-t._-._. ....t...... ~- I _-+. .....J..-..._.........---- --.t 1. -- k... t.. ~-. ...-..1......-. _..."._.,..u._~-~... - I. s u ... _..._.,_ _... g.... _.1 ... _. _.. ~. _ _ - .-....,.......j...,..........t..... ....,___1........ o ._...g,_.._......,........ +... . 7...:. s .4.. .. _....*; - : i " j O' a. .. L.... g.... _ _.. ;.,. .t. .t,..~ ~.b__,, ,.~_.....f' _-- t._ =t.._-*.._.4a_,, ,b . d

4. ~q. r ' ' a _. 3 g

.[ -.. a _..,.,.. _4 _,.. a _,._.L a,.= j H_. ; ; j x e 'l.

...... y,.._._..--,.__.

_4 p~7 ___..._. it L...\\.._.... i 4.. _ i ,.t.. ....t. 1.. _ 1. t.,. ... -,._. l. . i.+ i. .1. _ii

4.... Q.

......m - ;..l. i _f..L., F a. i .. m _._7._ , 4. 1, _. 1 - ~ ,. L. ..m -7 e. t.._ 10 72 hours i..... 2 n r rpr.. 7;t-- rn , q ~ ~-- r-mt i-t-U- r.i tm t-- ....c a -.;. N.- p 9 e .,._...} r.s q u - t. 4.-s.- --s..-.-- i n -t- - -.. : -- - 2.- l u...... - s..:. ..rj - e : y p: _... _. g. :..;. : -4,. t-- g d i y s . g. e 4 :- : p...H.- : ;.f.- i. q.- i.;-'. t

j. :.

.? s .r .......g. u. 5 -. -p.- - m..._. .. !.. =. .e.. c :... :4..r.......=.:.

4. i. %..t..:)

-t.

I.

. L.. : } :.....;. :. :. t..-. y" .) L.:. ..=s.. '~ t. . t....... :. _..... _.. i.- = =.: -. u.s..:....v. .._ :u uJ. s.. s...:_:. _ ~ ..... L._=._..:. ;....:...t.... ; J'... T.. t

s,..G..

T..... l g .I...i....__...:.. .....t.... _.......... _.. ..- _......i....._..-.... _ w.T I

g.,. -' +

1

4. _...

4 .i..., -.

9as...n

.m., . _ y.,s . g._ .a i.t r -. g ..~T._._.. j _... l. ..l.. ..[.. -_m..._ .4. r ,3 a t e. i;. .. 8 g,_ i -.+ .:t ri J 6..- u g. i 6 i:....t .. = r i.. _.. -..y,..-.. m... t.-..... !... m_ .... g. l _. e.:f_g _ _ta.. .._.__o.m u..- . i.. .sr-.... = g. _. a. a .. ar :. -o - :. q.s c. v 5 l 6- -w.-.. .._- u. ' Wr .. i. .. :, d W

) s'r l.. ! I i_.. l. ;'s l '. : " J O Ik2.s- :

.2.- P., -. f..? -Yi-. F O* i. l.- -.. - 1 .A..."............ C_:.... ...t......... %.. ~. _.t...:...i.._..... J : .._....~..t.u. ..'r....a. i :.. m :..... -..-__ .t 4 11 .t g ....l._.... .t....... ... r... l .t _...._...I.- 4 ..l....,._.. ......I._.._. ... _....,. _., _....... ~....... ....i.,...4 ......g. i.. \\..._...... 4 .t........ ... i............ . _.. g ....3....... 624 hours....,'.. g .}.....6...,.... .,__..,_.....]..._........l... ~.......i . w.....,.. %..,. l. .. + ... }.. 4. 4_ ,.. 1. _.. 3.... _ _ ..,9 -.r.... ...q9 -.-o. l.. ..g .h -.,..... l.... } 4 L.... ......w.. . ; 7. .s L.. ,. _... I. ...,..a. . i ..i.. .I -.g... t fe.. ..I.. ..e l ...6....,.._ . L..i _:. e , n ..( ... a :. ..q. ..t..6. -.l. .. l. .o . a..,1..... .._ e ~1,

m..

=8 i 1 . 4. l. _.:,...i .l. i i. m.. 7_a.

4..

10 s. 7 e- < \\ e m 6n u u o i a a i. n m a w u a w sa n u u mm n Probabilitv. 4 4.. a im e rn ai +m -r-. T .----7 m-i'V

Oystar Creek ?, ' Dirsction Dependent FIGURE 7 LPZ - Elevated Releases se e, se es = ao se .e in to se s , es e, e, en en i ~$" 10 .p-4 q.m i.i. +... #... ,,.3...i .i s. a :... n .;...I. ,9....4; - :l ..t.: t :. 4 :. :h 4. l.- i t --l

.i

',[- .: 1:.i : Jr. i. !:I. D ' I '.W M n _ l.. i. b i- .I-1

% :4 -. -

...l. ...i hni ' t. .. j.: ~ VIF:. _ n: - - !'"- l ii? @ Id Dii S I : ~ I :.J.5 3 :: IY i:- ' -- M.!- : :.

7:Ey -.+.

t i i, t.,.. ._i _ _}. . i_ .5

0

) 4 -.~... I.. t s_ L. t-gs, t .6 S-- I - i i 4...t_. ... ; b... a. pr ~. - - u- -q ..i.. : _ -M .i....t....; 4 a

- r....

p. i:.l. ..i. .c

4..

...it- -t..:. .ru. . i.:a... . :. 2...a.. =.. >i :. : w._:r.u. :.. _=._+

o b. ". :. i. N. p_:.: U.

i. t.

i ...g. --..--.1

l.. - :..:.2.. :. -::l-

n.:

e ........: :.-l.a.-.

...q =_=...:- g :. :. {. :.- I.. ;"..

c: ,~~ .=.r_n..u.rl..1._: :. l.=.._..,t ~....... p_. a._n : [.. :.. 'i ~.._ 1 2_:.. n..-...}...... .._.n.. 1 hour .=._=..: _.:. m_.= :._. t_. _._._.......~.....;---.......s_... -... _ -........ _... m 4 _.5, ...... w _ - 4......... ... ~........ 3. ,\\ _. J _..., _. t -.._.4..~.. -.....p....-.~-...~, 8 hours. {.. -. i ,. - -...,-.6 ~ + . ~ e.. I ..L.._~~J.....~_._4~,.....- .} . l -y . _.4 i -a i. ...... A. _. ..;.......p.

- g

....-..L._ ...,....t-- >+ .. *._. *I _ _l.... ...p. _L.i * - W '.. '[._.. I'.._

__ ;t'.1
n'....

__.u'. c

a 3

u. -6 1 ti 10 '..'... h . i.. t.. i - .. i. i. m u.4 i.:. t n.i. c.N .- Q. p...-... N * ::.l.\\..7.sv. t:: 5. - "- i.r.. p. .? ;. _. i s: )i- ..- i ;..

1. * :i.

s.. 6 ,[-. 5- > W. ?a "= i ..---i.- -t -....._..'... "; ! : :W: ! ' - M l.7-i.. 493. WWal :d-N;mN,r: f .-.......t ...-... _. 1..:_.=... r. :. ::. := : t = \\... -

= t.*. -

_ =..: : : ~ l =: - -- n. =.. :. :. =: r~g:. =_. ~.s..= t== :.L... ,M._....-... t. y. . _s -,.,

r. x

.. _.3.

4..

.,.t.. 2-"] -. ' ~ e ..-l..__ ..a._. ,...- \\.- a ..... e - b W --+ t -.. _ ai Ac i-~ ~ i,- i..... i at t. .. ~ . :- V.. '. 1......,..._ p+. c... n. 3.. 1_w s- ..=.c.:i-.p c. :i - :->, +.w. o. .. : L-a 4 -l -m ...- :. :.. :]..-.. g _.u_.; xagp t::-:g. u. t. -.,. r 2 e -.- .1 t-1. ,.s u -1 E--9. y.' My--c-7f-i-1 Uif N:d:i:g.L-Ti- : -! '-. !:'+- '!: :1 _

~

16 hours .=..-%. :,r- :::1_:- :. =.-

._. t. u..[-. _ :-.....:: ul.:. e a

_.:.4 =. :. -.--.-l. =..n. :. :. r. _.. e. c o .1 .. r._._...... i... g......,.-....:_... O. .___._,_.._.ft.._._..__-.___..._.,,....(_...,....._.i._..,.._.....i... _..__.J. __.t.___ -... _ L. \\... _. _..... _.... _.._.._._.t_t._.._. s, 1........_L......._,._.t... .__.__...._._t........ _.,s.... .... _ _.. _........ _.. _.., _g ,... _.., _...... e.. _ _..._..i..... _.._s_.; o. ...__t... s..- g. s_ .._._u_ __L.._..s. _.... N.. _ , =,.. __ L.... .i .._._....__.i.._ _, _.....a.... ...,_i.... _.t..........,. ..t.. _. %_. _....w - e p. .__ p.:. __j _.p...... _._4 .y .,.. y. . 1..__ _7 . \\. 10 72 hours - x. . ' - ~f r "i ;.

g. v9:.__

r----~- 4^ - + - " r - - u, [. g. i e r. r r:m y : -t

e-.$

' rycf.:. s.g. :. t 4_ 9. - ).. - {. :.N : :.:.: : t %;.. j-l. p. l :[ p {: ..l l ) .t. j t q.

i.. p m.

.t o :l [. l

j.
_: j :.~. t.

.. 3. g-n : :: ;N-[ :e i-

1 'l. ;l

. b- - l i h-I . i! E'il'.";[ j:i+Miki'l'.:-:I ;l' I: 11 -, Q.y l .l. q 1;_j_;..:_. q. q.__! _j..j, p._ '.-..j _'. g ._ :._ g... 4:: -i 4

l

1...N.

.L -l a. y..._ : . - - 1... 1 i F .I

t. '

.i ,,],,,,,_,;. (...

e..

-..r.... :: p. A: u g._i_,_;; '. j i b

V - [ '

' ~ h.'. ... 2 Q1 ..j b '" ]

i.

.. t. .:.. i 2 .~i-. '- ) .L-- l.,. h. *. -.,........... 8. l. ...,............l.... _............... .....) ......t_......a.. ... l. -... .q ..g.. .,..t,......._....4_. e ..p.. ,......_._...t

t... l.. 1

.t ..t. t. ,~ 624 hours.

a... i....

l __m..,.. - -.l _,,.,....l.... _2.___. . j...l.... ...).... ..:_~.. ... h.... $ g.... ..i.. s. t l..,. ...-_.g.. 4. _g ._ m I.r..1,,....l., ..e. ........e.. i._ 6* _8 +e 3. ) 10 .=.6.e...i _.....g I. e t l 49% 0.4% LA 8J 83 6 3 38 N 38 a8 34 W 79 ne to in 94 59 31 8 11 9 Probabilitv ..s p, -.,,,,,m-,.,-yj y. -v 9

Oyster Creek ..,,f Diratetion Independent FIFURE 8 LPI.- Elevated Releases \\ ,, a n = a a = a a i e i n . m o

5 ' i.

.,. gt... .., i .g i...i.i a ..w -.I. : i. .") n ! d .l. i-n -t. . p. l. 4..-6:

t v ' +.-% ' !N-l '

1.j

. t i :.. ... --t .I hour 1'- . : h:..

l-t -d.

D

-i .4 'I' L M. k I *' Z. - **.-

-: r. S \\'.... __ -.

g. E -* ! I 'I ..*\\ h

E /*

. L .t. l. T *1.1..*. ' * *

d M--U U. 7 *'.~- -

. b.+3. 5. -b f..N..* f_ *.

l. N...*!. TJ.* * -. k N.

._;m"Il..

E-I

\\. J [s. . ~ - - ..j.- ; ' g i --e. L k, ;. l 2 - 4_w* Li. r.. s.

. ' ~. t. -.. j.--. -

t.--. -- {.- -

4; - - e i ....tv 4 - 8-k..- 6.

a.a. -{.aj.

5 .:)

i a._,.._..

, g _. i..r-. t, ..s .I.. .g ... r... s - ..l. *-l

. 1*.

1-t T

4. . -..............g.. ...a... =.

. ;.1 -

....I. 1 .. _ _.... n. _.- -.. : '.:.. ~ : - '.. ' :..@' :.. -.. I. :.. 'l -- -. .... -.=-. 2 ^ ............... l -.. : - : l =. ~1 -. - -...]V ~. u

. \\

.: t - -..........-.~-. .~-..~t ~ - - -. d...... 1.. -..... .. ~ g ..,-.............-.--..~n... .. ~... - - .....l........_....,_t...... --e.- ......~.~..t.~ ..---.--o._.. ,,m......,-.,,.,...e..... . ~ - -.. *. + " ~ ~.. . - *...~-* -

~

.4 s-=**... = *. -. +' * *. = ~. s ..g.-._ o......6 ..,t...d. r.-- i. ._4 .i -....-i.. .4 l. a. .. -. + _ g.. - .-.1.. .< -.L.. t...< t. e .. - g== 6. . 1 8 hour, . -.. = ...i_ + -w - } - - t..- .-+- n4 r-- ,..-.s -. l - .... }. - - ..i 2.

_@ a

.,,2,.t.7. t .--' T' - P" .. '.a , N..... t. ,.. _t.. 't ~ ~- ' ' ~ ' " ~ ~ 10 T~ ^ i "s~ "~ "~ i -6 T "*"i

i M

i i 1 .ew, i.- m .. N.. : _o c. m N. i.u.b u.i ..n: y:i qp ..a.. Lr .-;..q... > p_t:.;,;:

m.. _.

.L g.. -.I... --. v ru. s... c M -> t'. " % i "-H ' '- = l 0M8 E-Ril i M~.:=-L-I6:.9=6# % ~ @Mi =- U W.: :1= 4. o .....:..:.....;...:.u .._ _.,...- c._._ t. _:... .\\..:s g.r. .,. v... t... :......t......

1..-

_. :.: _m....r.... :.. t=_ __n :. _r_....

.:c :_..

._._.g__ u.-_: . t. -. j .- L ; # e t s.% T '.L....,t ..t,,.. i 4. .1 - - . --. n t. 1_.. 4. .t.. e.. \\: I u.: .y ,.i.- %. s. w s,. i l-. s. : t '--~ ..,. n _:. .l

9.n.'.

p.a . l.... H. t. r.., r-

j S.-

l:. . : W-- -.~; i-. za ..._ --r. r ei u = pu. :.:v..= j ; q...;. - ~- -+r s.-- ^ (9 ..e..: i --;?-* ~ -" - Y I-h. N *3 - -.-.n.. M iE' Y i g 4. . b*-b-h

.Mh

- '. j.4 ' i 5?i .~t--'.. h.*.. b * .1.--- g.... t...j........... N .g .g...l.~.....,4,...1._. .....g .QQrS 1.,. g. u D 10 .. _.J. }......... ._.y......_._,,,... __....,...i ..t._~. g ..i .....i g4 g__ ..__..6.._,..!.......~!.........L......l......~..l........ 6... i._...._~_._..I...

t...

.,..........d.......,...g=...6.,.....- .._..I.......,,......_i........6..... CY .s.g.. ...... +.. . }.... p g .)..i.. 6 l _.. .M ..i.,..... ..) .,..j.6.......,..4.. + 2-._ _,.._..-J..-'. ..p... A 4. b. m .s .-_._. IN...; '. .... {. _.. . j.... ...t. 4. .... w -E -I l. 72 hours m. ... j.q :_ ,.. g.. - -...,,, t y.....N.... L ,....l h .l.. ..r 7. ,.o. L.. ..i ... x. 6.. .t.

e 7

o 4 ,. T.'" 10 I... 1 .m.- p. -I r :). .i...l...; t i..i. i.l

iI~j 3

.l. s-( ,.i t -- I-t ap.t .L. -} s.. e

g.

j. :. ...I .=. u.%. u t-: 1 t-p .t-

l.

[ ' t.. - -l-s. il. l.: ? -+

e

........f _..... t. . l. :. ........... l.. i t

p I

.g. l _. s.. _c. l : .i. 7 i L_- -. r, _ 4-r ~ -. ' _.... _ i l j. t. s. j. e t-e u A 6 l- .l . g g.. S <{. t ,..I. t 1.' 7,

g......? _1.

e J....

.4 . _.'L. -.1- .j. -i. -.- 5 l o l

*t-l :

I 3 .=.:: 5 d. I:

lM i.-..

L. J} 1. i.*.. 1:. ..t. I,

4....

e. i..--. 'f :

l-i.

.[ ~ L ri l t i ' ' i. i.:.; 1,..,.-. s-

a......;..,.

.i.........i.... . i. ......t. .t...........t.. ...t.....__... .l... .i.... i .......... L.. . l..... g.... ...,.._._.L...... .s... . ~..,...... r......... . g.. p. .g. s .. : _.. t..._... ... \\... 4 o..i ^- 624 hours :. . ~.... .p. . l.-....- + .....1.._.-.. .i _.. t . t.... . L. .g .: t ,t.. l. t. 10-g ... n. in .a.i u o i a s i. n m a w a 8 o n u I Probability w ,w.-- a w- ,wgw

TABLE 1 i Summary of Accident X/Q Values (s/m ) Elevated Releases Oyster Creek Meteorological Data j 7/01/76 - 6/30/77 j Averaging Time 0.5% Probable 5% Probable 0.5% Probable 50% Probable Direct. Dependent Direct. Independent Direct. Dependent, Direct. Independent Period After Accident (hours) EAB EAll Lpg Lpg (414 m) (414 m) (1208 m) (1208 m) 1 1.7 E-6 1.7 E-6 1.7 E-6 6.0 E-7 ,3.8 E-7 9.1 E-7 t 8 9.0 E-7 8.4 E-7 i 16 2.5 E-7 2.3 E-7 2.5 E-7 1.1.E-7 72 1.8 E-7 1.2 E-7 1.7 E-7 6.2 E-8 1 l 0 .624 2.5 E-8 2.5 E-8 2.5 E-8 2.5 E-8 l I ~ t l l; ), t ..- a e l

_...._x... TABLE 2 a 3 t I Summary of Accident X/O values (s/m ) Ground Releases Oyster Creek Meteorological Data 4 7/01/76 - 6/30/77 i Averaging Time 0.5% Probable 5% Probable 0.5% Probable 50% Probable Direct. Dependent Direct. Independent Direct. Dependent Direct. Independent Period After l Accident (hours) EAB EAB LPZ Lpg I (414 m) (414 m) (1208 m) (1208 m) i 1 7.7 E-4 6.3 E-4 1.8'E-4 4.0 E-5 t s 9.6 E-5 2.3 E-5 8 4.2 E-4 3.0 E-4 s A 16 .l.7 E-4 1.4 E-4 3.8 E-5 1.1 E-5 72 1.1 E-4 7.6 E-5 2.1 E-5 7.7 E-6 1 l_ t 5.2 E-5 4.3 E-5 1.0 E-5 5.7 E-6 624 i 8 1 p l 4 i

+ u~ ] 4&v, r, l SEP SAFETY TOPIC EVALUATION OYSTER CREEK NUCLEAR-POWER. STATION TOPIC II-4.F SETTLEMENT OF FOUNDATIONS AND BURIED EQUIPMENT ~ INTRODUCTIONE The purpose of'this safety topic-evaluation is to review the plant geotechnical engineering aspects.related to the properties and stability of subsurf ace materials and founda-tions.as they influence the static and seismically induced settlement of Category I structures and buried equipment. The scope of this assessment includes: a) review of the orig - inal design and construction information (if available)! f or .the Oyster Creek Nuclear Power Station; b) review of the re-sults of more recent investigations by Jersey Central Power and Light's (JCP&L) consultants, and c). evaluationEof the ef fects of L existing and future settlements on Category I.struc-tures using current United States Nuclear Regulatory Commis-sion (USNRC) review criteria. CURRENT REVIEW CRITERIA The evaluation presented herein is based on the regula-tory criteria and guidelines presented in the following: USNRC Standard Review Plan, Section 2.5.4, " Stability 1. of Subsurface Materials and Foundations"; 2. Code of Federal Regulations, Title 10, Part 100, Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants"; (Revised 30 September 1981)- 9

Prga 2 3. USNRC Regulatory Guide.l.132, " Site Investigations for Foundations of Nuclear Power Plants"; and 4. USNRC Regulatory Guide 1.138, " Laboratory Investiga-tions' of Soils' for' Engineering Analysis and Design of Nuclear Power Plants". RELATED SAFETY TOPICS AND INTERFACES Related SEP topics for the Oyster Creek Nuclear Power + ~ Station-include the following: 1. Topic II-4. A - Tectonic Province 2. Topic II-4.B - Proximity of Capable Tectonic Struc-tures in Plant Vicinity 3. Topic II-4.C - Historical Seismicity within 200 Miles .I of Plant These topics have been addressed in Reference 9. DESCRIPTION OF OYSTER CREEK SITE AND FACILITIES - ~ The Oyster Creek site is located on the Atlantic coastal plain of New Jersey. The site, which has an area of about 800 acres, is partly in Lacey and partly in Ocean-Townships of Ocean County, New Jersey. The site is approximately two miles inland from the shore of Barnegat Bay. The ground surface in the vicinity of the nuclear power station is at about el 23 ft II. At present, the Oyster Creek Nuclear Power Station consists of Unit No. 1. (1) Mean sea level, USC&G Survey (Revised 30 September 1981) r.

e j Page 3 The structural and foundation information about existing structures were obtained f rom drawings of as built conditions prepared by Burns and Roe, Inc. The seismic classification of existing struc:ures-an'd components-were based on informa-tion provided by GPU Nuclear (Reference 4). For the purpose of this review, we have evaluated the settlement conditions of the Category I structures described below: Foundation Bearing Maximum Bearing S tr uct ur e Elevation, ft Material Pressure, ksf Reactor Building -29.5 Natural soil 11.3 Vent Stack -10 Natural soil Not Available Intake Structure -18 to -21 Natural soil Not Available Discharge -12 to -13 Natural soil Not Available Structure Circulating Water -14 to -17. 25 Natural soil Not Available Tunnels Excavations for these Category I structures were made to approximately design foundation elevations. GEOLOGY AND SUBSURFACE CONDITIONS AT OYSTER CREEK The geology of the area is described in detail in Refer-ences 6 and 10. The Oyster Creek site lies on a thick prism of coastal plain sediments. The ages of the sediments vary from Cretaceous to Recent. The bedrock in the area is early Paleozoic or Precambrian in age, and is estimated to be at a depth of about 3000 ft. The sediments were deposited under relatively stable shelf conditions and were laid down in continental, transitional and marine environnments. (Revised 30 September 1981) -r - - - ~

Page'4 The subsurface conditions.within, or in proximity to, the site of the Oyster Creek Nuclear Power Station were investigated in five major subsurf ace exploration programs as described below: 4 Subsurface Year of Exploration Program Exploration Ref erence No. General Site Area 1960 5 Oyster Creek Unit-1 1964 1,7 Oyster. Creek Unit-2 (propos ed) 1968 2,3 Forked River Unit-1 (proposed at a 1970 6 location about 3000 ft west of Oyster Creek Unit-1) New Radwaste and Of f-gas 1973 10 to Buildings at Oyster Creek 1974 Unit-1 All of these investigations indicated that.the subsurface geology at Oyster Creek was mainly uniform, and the existing subsoil strata underlying the site extended more or less hori-zontally and had similar geotechnical properties. The specific formations encountered at the site consisted of five soil strata. A detailed description of each stratum is provided in References 2 and 10. A summary of the forma-tions encountered at the locations of the Oyster Creek Unit-1 structures is given below: (Revised 30 September 1981) ~

~Page.5 Average Thickness Description of Major Formation and Elevation, ft Soil Component Cape May 17 (el-23 to el 6) Tan medium to fine SAND Upper Clay. 17 (el 6 to el -11) Gray organic CLAY, with inclusions of' fine sand Cohansey 65 (el -11 to el -76) Yellow-brown, or tan, medium to-fine SAND Lower Clay 8 (el -76 to el -84) Gray medium to fine SAND with inclusions of gray organic clay ~ ~ Kirkwood below el -84 Gray medium to fine SAND The groundwater levels were found to be less than 10 ft below ground surface during the 1964 subsurface exploration. The 1973-1974 subsurf ace exploration indicated that the ground-water level in the Cape May Formation was perched at approxi-mately el 12 ft, whereas the piezometric level in the Cohansey Formation was at approximetely el 6 f t._ EVALUATION 1. Bearing Stratum far Categorv I Structures: The major Category I structures in the Oyster Creek Nuclear Power Station are supported by mat foundations bearing in the Cohansey Formation. i l The static and dynamic geotechnical properties of the Cohansey sand have been investigated extensively by standard penetration tests, index property tests, relative _ density determinations on undisturbed samples (1968, 1970 and 1973-1974 investigations), cyclic triaxial tests on undisturbed s amples (1973-1974 investigation) and by a seismic velocity survey (1970 investigation). The test results are described in detail in References 3, 6, 10 and 11. (Revised 30 September 1981)

Page'6 All available information and test results confirmed that the Cohansey sand had a dense to very dense relative density. These results also indicated a marked increase in _ standard penetration resistance (N-values) at about el -30 ft. The direct determination of relative density f rom undis- ~ turbe'd samples indicated that the relative densities of the Cohansey sand were greater than 70% (Ref erences 3 and 10). Based on the 1964 test borings logs, the N-values mea-sured-in the upper Cohansey sand (above el -30 ft) had an average of 55 blows /ft. In the lower Cohansey sand (below el -30 ft), the measured N-values had an average in excess of 100 blows /ft. In general, the N-values measure'd in the Cohansey sand in the 1964 test borings were substantially higher than those measured in the more recent 1973-1974 test borings. However, we should also indicate that the two series ^ of test borings were made by two different boring contractors. 2. Settlement Under Foundation Loads The major component of the expected total foundation settlement of Category I structures is due mainly to the com-pression of the dense Cohansey sand. This settlement would have been virtually completed by the end of construction. A plate load test was performed at el -30 f t during con-struction of the reactor building and resulted in an observed settlement of 0.103 in under a test pressure of 20 ksf, thus confirming the very low compressibility of the Cohansey sand (Ref erence 6). Soil consultants Casagrande and Casagrande estimated settlements of "less than one inch" for the reactor (Revised 30 Septe:ber 1981)

Page 7 building (Ref erence 2). Observed settlements of the reactor building f rom the start of construction until May 1968 have ranged between 2/3 in and 3/4 in. (Ref erence 2). Therefore, _ we can conclude that f or practical purposes, the anticipated settlement of Category I structures founded in Cohansey sand have already occurred. Any additional total and dif f erential settlements in the future are. expected to be negligible. 3. Licuefaction Potential In the 1968 investigation for the proposed Oyster Creek Unit-2, soil consultants Casagrande and Casagrande-concluded that cohesionless soil below el -23 ft at Oyster Creek would not liquefy during a "very severe earthquake" (Reference 3). - Their - conclusion was based on the high N-values and relative densities measured on undisturbed sanples below that eleva-tion. In the 1973-1974 investigation for the new radwaste build-ing, cohesionless soils below el -30 f t were considered not liquefiable. The liquef action potential ce sand above el ~ ~ -30 f t was evaluated by two methods; i.e., the simplified and procedure developed by Seed and Idriss (Ref erence 8), a method in which a direct comparison is made of the stresses that would be induced by earthquake motions with the cyclic strength characteristics of the soils. The induced stresses were estimated by a seismic response analysis, and the cyclic strength characteristics by laboratory cyclic triaxial tests. The two analysis indicated that the sand above el -30 f t would not liquefy during the postulated Safe Shutdown Earthquake (SSE) or the Operating Basis Earthquake (OBE) with peak ground accelerations of 0.22 g and 0.11 g, respectively (References -10 and 11). (Revised 30 September 1981) ~

l Pnga 8 The currently accepted SSE peak ground acceleration fat Oyster Creek Unit-1 is 0.165 g or 75% of the SSE peak groum$ acceleration considered in past studies. Our evaluation of the subsurf ace inf ormation leads us to conclude that subsur-f ace conditions at' the location' of Unit ~-1 are similar to ticse-at the location of the new radwaste building. Therefore, we conclude that the Cohansey sands under Category I str uctur e of Unit-1 are non-liquefiable during the SSE or OBE events. CONCLUS IONS Based on the inf ormation provided in the available ref er-ences and the evaluations ref erred. to above, we conclude that the original geotechnical design bases which influence the static and seismically' induced settlements of Category I struc-4 tures of the Oyster Creek Nuclear Power Station are compatible with current USNRC guidelines as described in Standard Review Section 2.5.4. Specific conclusions are listed below: 1. Future static total and differential settlements of Category I structures and buried equipment will be very small and should not create a saf ety problem; and 2. The Cohansey sands are non-liquefiable under a Imstu-lated SSE earthquake with a peak ground acceleration of 0.165 g. 1 (Revised 30 September 1981) -r-- = v -,n y

Page 9 REFERENCES 1. Burns and Roe, Inc. " Plan of Test Borings Made by Sprague and Henwood, Inc. for Jersey Central Power and Light, Co.", Sheet 2 of 2, 1964. (The boring logs are also included in the PSAR f or Oyster Creek Unit-2, document not available for this review). 2. Casagrande A. and Casagrande, L. " Report to Burns and Roe on Foundation Investigations for-the Oyster Creek Nuclear power Plant, Unit No. 2", repor t to Burns and Roe, Inc., 1968. 3. Casagrande A. and Casagrande, L. " Oyster Creek No. 2, Relative Density Tests on Undisturbed Sand Samples f rom Borings 19 and 20", letter to Burns and Roe, Inc., 1968. i 4. GPU Nuclear, " Oyster Creek Component / Subsystem Seismic Category Groups", in a letter to USNRC, 1979. 5. Jersey Central Power and Light Company, " Oyster Creek Nuclear Power Plant Preliminary Saf ety Analysis Report", Section VI, 1960. 6. Jersey Central Power and Light Company, " Forked River Nuclear Station Unit-1 Preliminary Saf ety Analysis Re-port Vol. 1", 1972. 7. Jersey Central Power and Light Company, " Oyster Creek Nuclear Power Plant Unit No.1 Facility Description and Saf ety Analysis Report Vol.1",1967. 8. Seed, H.B. and Idriss, I.M., "A Simplified Procedure for-Evaluating Soil Liquef action Potential", JSMFD, ASCE, 97, No. SM9, 1971. 9. URS/Blume, " Seismology and Geology Oyster Creek Nuclear-Generating Station",1981, 10. Woodward-Clyde Consultants, "Geotechnical S tudy Proposed Radwaste and Off-Gas Buildings Oyster Creek Nuclear Power S tation", report to GPU Service Corporation, 1975. 11. Woodward-Clyde Consultants, " Response to NRC Questions' Radwaste Modifications Oyster Creek Nuclear Power Station", 1975. 4 (Revised 30 September 1981)

e 2..- + 'IUPIC V-12A WATER PURITY OF BOILING WATER REACTOR PRIMARY COOLANT 1 I. IntrBduction This topic reviews the primary water monitoring and reactor water cleanup - system capabilities, including the water purity, to determine if the maintenance of the necessary purity levels comply with the current criteria. Evaluation of the Oyster Creek Nuclear Generating Station primary coolant chemistry control with respect to current pertinent regulatory criteria is provided. Any deviations from current licensing requirements are identified and the significance of the deviations are explained. II. Basis for Evaluation The current criteria pertinent to this topic is: U.S. Nuclear Regulatory Commission Regulatory Guide 1.56 Revision 1 of July 1978 III. Evaluation The following is a breakdown of the compliance of Oyster Creek Nuclear Generating Station with the criteria set forth in regulatory guide 1.56, maintenance of water purity in boiling water reactors. Regulatory Position I "The licensee should establish appropriate limits for the electrical co.ductivity of purified condensate to the reactor vessel (the electrical conductivity of the BWR feedwater cycle and that of the reactor water cleanup cycle). Separate limits may be required for such operating conditions as startup, hot standby, low power, high power, and at temperatures $212 *F (100*C). Chemical analyses for dissolved and suspended impurities should be performed as called for in the plant technical specifications. A conductivity meter should be provided at each condenser hotwell or in the time between the hot-well and the condensate demineralizer with sufficient range to measure at 1 cast all levels of conductivity up to and including the limiting conditions of the technical specifications that require immediate shutdown of the reactor. The recording conductivity meters recommended in regulatory position 4. A may be used for this purpose." Reponse Oyster Creek Technical Specification 3.3E includes the following limits on primary coolant: _ _,. - --. - -, ~... - -. c.- ,n

t Regulatory Position II "The licensee should establish the sequential resin regeneration frequency or resin replacement frequency required to maintain adequate capacity margin in the condensate treatment system for postulated condenser cooling water inleakage. The capacity required and corresponding resin regeneration of replacement frequency will depend on several parameters, including condenser cooling water composition, chloride concentration, flow rate in each demineralizer unit type and quantity of resin, cation / anion resin ratio, postulated condenser leakage, and time for orderly reactor shutdown."

Response

Regenerations are performed as necessary to maintain acceptable water quality. The original General Electric Company recommendation was to regenerate one bed every three days. Based on past operating history, regenerating one bed per week is an adequate frequency. The selection of the demineralizer bed to be regenerated is made on a rotating basis unless unusual conditions dictate otherwise. Regulatory Position III "The initial total capacity of the new anion and cation demineralizer resins should be measured. Anion exchange capacity may be determined by a procedure recommended by the resin manufacturer. The total exchange. capacity of the cation resin may be measured by a procedure recommended by the resin manu-facturer or by paragraphs 41 through 49 of ASTM D2187-71. " Standard Methods of Test of Physical and Chemical Properties of Ion-Exchange Resins." For resins that are to be regenerated, these determinations should be repeated at least semiannually. The resins should be discarded and replaced when their capacity following regeneration falls below 60 per cent of the initial value. More frequent determinations should be made at plants using seawater or other water containing large amounts of dissolved or suspended matter as coolant in their heat exchangers. For resins that are not regenerated but are instead replaced periodically with material of the same type, measurements of initial capacity should be made on a sample of new material at least once a year (when the time between replacements is less than 1 year) or at each replace-ment (when the time between replacements exceeds 1 year). When the type of anion or cation resin is changed, a measurement of total capacity of the replacement resin should be made prior to use in the deminerali:er."

Response

Oyster Creek procures all of its resin under a rigid purchase specification No. 34 4. In additien, the vendor must send representative samples of each batch of resin for analysis by GPU before shipment is authori:ed. Condensate demineralizer resin is sampled after every regeneration. The anion salt splitting capacity must be f 60', and the cation salt splitting capacity must be ? 70'6 of that of new resin or the resin must be regenerated or replaced. . ~. ~.- . - - ~ ~ ~

1 .3 Regulatory Position IV "The licensee should verify that the minimum residual deminerali:er ~ capacity in the most depleted demineralizer unit established in accord-ance with the recommendations of regulatory position 2 is maintained. The following is an example of an acceptable method for determining - the = condition of the deminerali:cr units so that the:ica exchange resin- .can be replaced or regenerated before an unacceptable level of depletion is reached. A. Recording conductivity meters should be installed at the inlet and outlet of both the condensate treatment system and reactor water cleanup systen. The range of these instruments should be sufficient to measure all levels of potential water conduc-- tivity specified in the plant technical specifications. For the condensate treatment system, the conductivity meter readings should be calibrated so that estimates of condenser leakage can be made based on cooling water conductivity, condensate conduc-tivity, and flow rate. The chemical composition of the cooling water and its relation'to specific conductance should be estab-lished and periodically confirmed so that estimates of residual demineralizer capacity can be made. B. A recording flowmeter should be used to measure the rate of flow through each demineralizer. C. The quantity of the principal ion (s) likely to cause demineralizer breakthrough should be calculated by: 1. Converting the conductivity readings of the water entering the deminerali:er to weight fraction (e.g., ppm cr ppb) of the principal ion (s) and 2. Integrating over time the product of concentration of this ion (s) and demineralizer flow. The input quantity of ion (s) to the demineralizers shculd be determined at a frequency adequate to ensure sufficient residual ion exchange capacity in the event of a major condenser leakage to prevent exceeding reactor coolant limits. D. Each demineralizer unit should be replaced or regenerated when l the remaining capacity (calculated by subtracting the total utili:ation determined from conductivity and flow measurements ( in accordance with regulatory position 4.c from the ' initial capacity determined in accordance with regulatory position 3) approaches the minimum residual deminerali:er capacity deter-mined in accordance with regulatory position 2. The accuracy of the above calculation should be checked by measurements made on resin samples taken when dcmineralizer units are removed from service for regeneration or resin cleaning. v m ~w

u --z -Measurements on samples from each unit should be made at each of the.first two such removals from service and dt every fifth such removal from service thereafter. If appropriate, the actual measurements may be used to adjust the calculated value of residual demineralizer capacity. Such adjustment and its justification should be reported to the NRC in the annual ' operating report." -Response In addition to the conductivity meters discussed under the response to - regulatory position I, Oyster Creek has the following recording conductivity meters: 1. Condensate demineralizer effluent - one per unit - ~ 0-10 unho/cm . alarm set at 0.15 umho/cm. 2. Condensate demineralizer combined effluent - two meters 1, 1-10 umho/cm - alarm set at 0.25 umho/cm. 3. Reactor water conductivity - two meters 10 l umho/cm alarms set at 1.0 and 2.0 umho/cm. i l 4. Cican Up System effluent - one meter 1 umho/cm - alarm set at 0.2 umho/cm. Jersey Central Power 6 Light Company Procedure Number 523, Condenser Tube Leakage; and procedure Number 501, Annunciators and-Alarms; specifically cover actions to be taken based on conductivity measurements to maintain water quality. Past operating history has shown that acceptabic water quality is maintained using these procedures during condenser tube Icakage. Regulatory Position V "The conductivity ueter(s) located at the inlet and out of the deminerali:cr(s) [ of the condensate treatment system and the reactor water cleanup system should be set to trigger alarms in the control room when, as a minimum, either of the following ccnductivity levels is reached (values of which should be determined by the licensee): A. The level that indicates marginal performance of the demin-eralizer systems. B. The level that indicates not' ceable breakthrough of one or i more dominerali:ers." t

Response

i Oyster Creek meets this criteria as per responses for regulatory positions i I and IV. t l l l I i l e L

J T r ; ', Regulatory' Position VI "The chloride content in the reactor vessel water should be maintained as low as practic'a1. The ionic equilibria-of the reactor water should be controlled to ensure.a neutral pH. The licensee should establish limits for conductivity, pH and chloride in the reactor vessel water and should specify procedures to-be used-for their determination.. Acceptable reactor ~ water' chemistry. limits, are 'given in Table 1 of the appendix to this guide. If the limiting values of the conductivity, pH or chloride content are -exceeded, appropriate corrective actions as defined in the plant technical specifications should be taken."

Response

1 l The limits for chloride and conductivity as outlined in our response to regulatory position I compare favorably with the limits set forth in Table 1 of Regulatory Guide -1.56. pH is maintained within the limits . specified in the fuel warranty for Oyster Creek. These limits compare favorably with the limits set forth in Table 1 of Regulatory Guide. l.56. 1 IV. Conclusion 4 Based on review of the primary water monitoring and reactor water cleanup system capabilities and procedures, Oyster Creek complies with the intent 4 of ~U.S. Nuclear Regulatory Commission Regulatory Guide 1.56 Revision 1 of July 1973. 3 e k i f f o 1 - +

~ ~. ~~ ~ . ~,., _,. A. steaming rates <100,050 - _1b/hr. conductivity 925'C 2 umhos/cm chloride 0.1 ppm B. steaming rates >100,000- Ib/hr . conductivity 925'C 10 umhos/cm chloride 1.0-ppm-. C. If the above criteria cannot be met, 'he reactor must be placed in cold - - ~" shutdown. t Oyster Creek Series 800 procedures include the following limits on Clean Up System Effluent and Feedwater Quality. If these limits are exceeded, operations personnel are notified. Clean Up System: conductivity.025*C .0.15 umhos/cm pH 5.5 - 8.0 silica (ppm) 1.0 chloride-(ppm) 0.05 Feedwater: conductivity 025'C 4 standard 0.1 unho/cm off standard limits 1.0 umho/cm Total Metallic Impurities standard limit 50 ppb off standard limit 100 ppb Dissolved Oxygen standard limit 50 ppb off standard limit 100 ppb Conductivity meters are provided at each hotwell. They have a' range.to l 10 umho/cm with the alarms set at 0.3 umho/ca. There are three conductivity meters on the common condensate polisher -influent. These have ranges 1,10, and 100 umho/cm. The Hi alarm is set at_0.3 umho/cm and the Hi Hi at 50 umhos/cm. All of these parameters are recorded in the control room. 4 In addition to chloride and conductivity measurements, suspended solids, silica and pH analyses are done routinely on the reactor water and serve as indications of how well the condensate deminerali::ers and clean up system are performing. .,u,-_. ,w. e*N T P

~ 4 , -.1* f.4* '.S-i t 1 SEP. SAFETY EVALUATION 0F RADIOLOGICAL CONSEQUENCES OF FAILUP2. 1 J OF SMALL LINES CARRYING PRIYARY COOLANT OUTSIDE CONTAIMENT l1 TOPIC XV-16 ,4 i FOR 1 i OYSTER CREEK NUCLEAR GENER/ STING STATION a ~ 1 ~ J 4 4 1 I 7

=

4 i I a l a j j J t 4 4 i [. l,

. y-1.0 L INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic XV-16, " Radiological Consequences. of Failure of Scall Lines Carrying Primary Coolant Outside Containment," is to assure that any release of radioactivity to the environment as a result of Postulated breaks in these lines outside containment will be'sub-stantially below the-guidelines of 10CRF100. i 2.0 REVIEW CRITERIA i The current criteria for review of radiological consequences j ~ of failure of small lines outside containment are contained in Standard Review Plan (SRP) 15.6.2, " Radiological Consequences of Failure of Small Lines Carryi5g Primary C6olant Outside Contain ~ ~ ment" and by reference,' Regulatory Guide 1.11, 10CFR Part 100 and ~ 10CFR50, Appendix A, General Design Criterion (CDC) 55, " Reactor Coolant Pressure Boundary Penetrating Containment." In the case of'0yster Creek, all of the small lines carrying pri= arf ~ coolant ~ ~ ~ outside containment, with the exception of a sample line, are instrument lines which are exempt from the requirements of GDC55. These lines are reviewed against the requirements of SRP 15.6.2 i and Regulatory Guide 1.11. i v f J 4 P .,-g ,r --e w 1m -r ga, y -9 ,,myq

3.0 EVALUATION-e

3.1 DESCRIPTION

OF SMALL LINES PENETRATING ThE OYSTER CREEK ' CONTAINMENT There are 59 instrument lines at Oyster Creek which extend from the reactor vessel or primary system through the primary containment to instru-ments and gages in the reactor building. Thirty-one (31) of these arc associated with sensors for the reactor protective system while the remain-ing twenty-eight supply indication and control-instruments. Each line con-tains two isolation valves, a manually operated stop valve and an automatically operated excess flow check valve, located as close to the outside of the dry-well as possible. A typical installation is illustrated schematically in Figure 3-1 and the instrument lines are tabulated in Table 3-1 with details of. their use and installation. The piping from the reactor coolant system to the excess flow check valve (EFCV) is one inch schedule 80 stainless steel while downstream of the EFCV 1/2 inch schedule 80 piping is used. A minimum amount of one inch piping is exposed outside of the primary containment since the isolation valves are located as close to the penetration as is possible (see Table 3-1 for the individual distances). In no case is the automatically actuated valve more than 32 inches from the contain=cnt. The excess flow check valve is a simpic design which cuts off flow completely should the outward flow execed 2 CPM. It is re-opened automati-cally when pressure is equalized across the disc indicating that dcunstream integrity has been restored. The 2 CPM setpoint has considerable built-in margin since the flow resulting from a 1/2" line which has been sheared of f is on the order of 50 CPM. 1 _1-

i TABLE 3-1 .4 OYSTER CREEX NUCLEAR GENERATING STATION INSTRUMENT LINES PE'NETRATING THE PRIMARY CONTAINME"4T .i s INSTRUMINT PROTECTIVE DISTANCE (INCL!ES) ISOLABILITY

INSTRUMENTS OR CAGES LINE SYSTEM BET'JEEN CONTAINTIENT OF ENTIRE ON LINE (MA'."JAL VALVE NO.)

(YES/NO) AND LINE IST VALVE EFCV* (YES/NO) SYSTEM P..V. bide range level ind. V-139-15 No 6 11 No V ' 39-95 No 11 19 No 1:.v..iide range level ind. I V *30-13

Yes, 6

11 No Lov, lei, low level for auto. g depress, actuation (RPS) I R.V. level for feedwater control V-130-14 No 6 11 *, , No

R.V.

level for feedwater control Lov,' low, low level for auto. V-130-2 2 B Yes 14 21 No i depress actuation (RPS) + Pressure for relief, valve actuation (RPS) l (See Note) V-130- 11 Yes 26 32 RNo High pressure for scram (RPS) High pressure for relief valve actuation (RPS) 2 l Reactor pressure bypass for vacuum, anticipatory and MSLlV 1 closed scrams ( RPS) I liigh pressure for emerg. coad. actuation.(RPS) i ,1 Reactor pressure-core spray y valve permissive (RPS) 1 p Note: Line from V-130-228 also provides pressure signal for fuel i zone level instrumentation and recirculation pump trip logic. j~, -2 () u

( TABT.E 3-1 OYSTER CREEK NUCLEAR GENERATING STATION INSTRUMENT LINES PENETRATING THE PRIMARY CONTAINMENT i INSTRUMENT ' PROTECTIVE DISTANCZ (INCHES) ISOLABILITY INSTRUMENTS OR CAGES LINE SYSTEM BETk'EEN CONTAIN:1ENT OF ENTIRE ON LINE (MANUAL VALVE NO.) (YES/NO) AND LINE IST VAI.VE EFCV* (YES/NO) SYSTEM J v-130-12A Yes 3 11 No R.V. 109, low level for actuation of core spray, cont, spray, d ry-- .: ell isolation, reactor isolation l and recire. pump trip (RPS). I R.V. 1.ou, lou lev 1 scram '(RPS) Same as V-130-12A i V-130-12B .Yes 6 13 INo Same as V-130-12A i V-130-16A Yes 5 26 .No Same es V-130-12A V-130-16B 9 Same as V-130-11 plus V-130-17 Yes 5 18 lNo Pressure indication ' t R.V. level for feedwater control g Lou, lov, low level for auto., depress, actuation (RPS) R.V. level for feedwater control No 5 2,4* lNo V_130-18 Low, low, low level for auto. V-130-22A Yes 5 22 ,No depress. actuation (RPS) 1 (See Note) Note: Line'from V-130-22A also provides pressure signal for recirculation pump trip. r q

~fALLE T h OYSTER CREEK NUCLEAR GENERATING STATION INSTRUMENT LINES PENETRATING Ti!E PRIMARY CONTAINMENT INSTRUMENT PROTECTIVE DISTANCE (INCL!ES) ISOLABILITY INSTRUMENTS OR CAGES LINE SYSTE:t BET 1!EEN CONTAIt;:11 IT OF ENTIRE ON LINE (YJ2:UAL VALVE NO.) (YES/:;0) A"D LINE IST VALVE EFCVa (YES/NO) SYSTEM ' 9A Yes 24 30 No - Iligh flow in main steam line for scram and isolation (RPS) Steam fle.r for feedwater control Steam line pressure in feed-water control Same as 9A 10A Yes 24 30 No Same,as 9 A 9B Yes 24 30 No Same as 9A 10B Yes 24 30 No 2A No 8 16 Yes Loop "A" Recire. Recire, pump AP indication 2B No 8 16 tes Recire. Recdre. pump AP indication Loop "A" 2C No 8 16 Yes Recire. Recire, pump AP indication Loop "B" 2D No 8 16 Yes Recire. Recirc. pump'AP indication Loop "B" 2E No 12 24 Yes Recire. Recire. pump AP indication Loop "C" 2F No 12 24 Yes Recire. Recirc. pump AP indication Loop "C" t' v

~ ~ i TABLE 3-1 OYSTER CREEK NUCLEAR GENERATING STATION I INSTRUMENT LINES PENETRATING THE PRIMARY CONTAINMENT j l INSTRUMENT PROTECTIVE DISTANCE (INCllES) ISOLABILITY' INSTRUMENTS OR GAGES LINE SYSTEM BETtJEEN CONTAINnE'4T OF ENTIRE ON LINE (MANUAL VALVE NO.) (YES/NO) AND LINE IST VALVE EFCV* (YES/NO) SYSTEM t 2G No 12 24 Yes Recire. ,Recire., punip,. AP indication Loop "D" 2 11 No 12 24 Yes Recire. Recire. pur.p AP indication Loop "3" 6 12 Yes ~ Recirc. Recire, pump,AP indication 2J No ( Loop "0" 2K No 6 12 Yes Recire. Recire. pump AP indication "l [ Loop "e"' Recire loop flow for power /, lA Yes 3 12 No flow scram RPS IB Yes 3 12 No Recire. loop flow for power / flow scram RPS i l 2ecirc. loop flow for power / 12 No IC Yes 3 i flow scram RPS tbo Recire. loop flow for power / ID Yes 13 12 flow scram RPS ~ a t-Occire. loop flow for power / ) IE Yes 13 20 No flow scram RPS Recire. Ibop flow for power / IF Yes 13 20 No flow scram RPS Recire. loop flow for power / ~ 10 Yes 8 18 No flow scram RPS i ~ -

~ r l t-l TABLE 3-1 l i OYSTER CREEK NUCLEAR CENERATING STATION INSTRUMENT LINES PENETRATING THE PRIMARY CONTAINMENT l 1 INSTRUMENT PROTECTIVE DISTANCE (INCL!ES) ISOLABILITY INSTRUMENTS OR CACES LINE SYSTEM BETWEEN CONTAINMI'NT OF ENTIRE ON LINE g (MMCAL VALVT NO ) (YES/NO) M;D LINE 1ST VALVE EFCV* (YES/NO) SYSTEM Recire: loop flow for power / i - 111 Yes 8 18 No flow scram RPS t Recire. loop flow for power / IJ Yes,. 8 18 No flow scram RPS i Recire. loop flov 'for power /' 1K Yes 8 18 No. flow scram RPS V-130-19 No 12 19 No Core,AP for poison system / reactor pressure for control rod drive system / core spray system AP I (See Note) V-130-20 ,No 12 19 No -i Core AP for poison s,ystem 29A No '19 26 No Core spray system AP 29B No 19 26 No Core spray system AP - 18A No 6 12 Yes Recire. - Recirc. pump seal cavity pressure g Loop "A" I 18B

No 6

12 Yes Recire. - Recire. pump seal cavity pressure i Loop "A" i i 1 18C No 6 12 Yes Recire. Recire. pump seal cavity pressure j I Loop "3" IBD No 6 .12 Yes Recire. I;ccirc. pump seal cavity pressure Loop "B" 1 Note: Line from valve V-130-19 also provides pre'ssure signal for fuel zone level instrumentation. I i C U O l

i o I ~ TABLE 3-1 OYSTER CREEK NUCLEAR GENERATING STATION ~ INSTRUMENT LINES PENETP). TING Tile P9IMARY CONTAINMENT i. INSTRUMENT PROTECTIVE DISTANCE (INCllES) ISOLABILITY INSTRUMENTS OR CAGES LINE SYSTEM BETWEEN CONTAINMENT OF ENTIRE ON LINE (MANUAI. VALVE NO.) (YES/NO) AND LINE-IST VALVE EFCV* (YES/MO) SYSTEM 6 12 Yes Recirc. Recirc. puc:p seal cavity pressure 18E No ' Loop "C" 18F No 6 12 Yes

Recire.

Recire. pump seal cavity pressure Loop "C" d 18G No 6 12 Yes Recire. Recire. pump seal cavity pressure Loop "3" s 1811 No 6 12 .Yes Recire. Recire. pump seal cavity pressure Loop "D. i 18J " No 6 12 Yes

Recire.

Recire. pump seal cavity pressure Loop,"E" 18K No 6 12 Yes Recire. 'Recire. pump seal ca'vity pressure Loop "E" 30A Yes 1 11 No liigh steam ficv in emerg, cond, to trip emerg. cond. (R?S) N 31A Yes 1 11 .o 1h No 303 Yes 6 i j 31B Yes 11 19 No 32A. Yes 6 13 Yes Emerg. Cond. High cond flow in emerg. cond. to trip emerg condenser' 39A Yes 6 13 Yes i u

TABLE 3-1 OYSTEP. CREEK NUCLEAR GENERATING STATION INSTRUMENT LINES PENETRATING THE PRIMARY CONTAINMENT INSTRt2ENT PROTECTIVE DISTANCE (INCllES) ISOLABILITY INSTRUMENTS OR CAGES LINE SYSTEM BETWEEN CONTAIN!!ENT OF ENTIRE. ON LINE (MANUAL VALVE NO.) (YES/NO) AND I.INE IST VALVE EFCV* (YES/NO) SYSTEM '32 B Yes' 6 13 Yes Emerg. Cond. liigh cond. flow in emerg. cond. to trip emerg. condenser 1 3')B Yes 6 13 Yes t I e I 9 1 e i' i t s 9 e e ? I i-4 9 '? I O ,m O I - e e l 4 t. ~ i 1 -8' l t i a 4 f. [

I PRlW A RY CONTAINHENT N)II [-l$a?'$' REACTOR t> 8 1/\\ VESSEL i*4.@

h. f',?."

h -J PRIM ARY 'J*.'3 CI C0tJTAlf4 MENT. ^ py dw.Yk,L*'.*I. PENETR ATl0N

U,h,.),'jj.'

lNSTRUMENT ,u. + v,. c .= Y.%,'0'.' (hh$j;fh EXCESS FLOW ,ttdQI

0 CHECK VALVE INSTRUMENT Id[Q (CESIGl4ED IS O L ATION TO SHUT AT VALVE y' *S1gKf,f I

ISOL ATION 2 G.P.M. FLOW) / VA LVE l DaC 5.P' 55 s..., W hi

IQ;}su.';.

ANG n.$h. EY -c C?$$I7 LINE I" SCH. 80 LINE 1/2"SCH. 80 .' i f;4 . ? DR AIN VALV E I,?. w.?l?:,', 30I:e :'

2*

U bb'd.E!-[.,: n CAP 's8,s:e vI,h MMf-] &. !lll:,1,': OYSTER CREEK NUCLEAR GENERATING STATION TYPICAL IN ST R U ME NT LINE FIGURE 3-1 ..c-

c.,

---4.,

r The' instrument line penetrations and~ isolation valves have been ~ ~ t examined to determine their vulnerability to normal industrial hazards such as vehicles, falling objects, rigging areas, etc. Most of the penetrations are located in relatively inaccessible places, i.e. 20 to 30 feet above-the floor, behind instrument racks or in rooms not accessibic .to mobile equipment'. None of the instrument line penetrations are located' ' ~ ~ ~ ~ under cranes or in direct path of mobile dolly's, fork lifts, etc. Quality materials were used in fabricating these lines and all welds - l were visually inspected, dye penetrant inspected and tested hydrostatically to th'c pressure of the system in which they were installed. These materials are listed in Tabic 3-2. This material was of the highest quality avail-abic at the time of design and construction and is consistent with the - material specifications of the larger diameter high pressure nuclear systc=s. ) In addition to the instrument lines described above, a 3/4" sample line from the "A" recirculation loop penetrates containment. Unlike the instrument lines, this line includes pneumatically operated isolation valves (V-24-29 and 7-24-30) both inside and outside the containment which are automatically isolated by the reactor protection system. These lines meet the requirements of GDCSS and 56. [ ~ 3.2 Operating Experience During over ten years of Oyster Creek Station operation, the instrument lines and their isolation valves have never cresented nrohlems. ] There has been no case where an excess flow check valve has closed spuriously. On one occasion, when an excess flow check valve was called upon to isolate a broken line, it stopped the flow completely after releasing less than one cup of water. This incident occurred when a fitting on an instrument failed as it was being tightened after it was noticed that it was dripping. 9- ~, ... x ,,. -,,,, m mum-c - s, y

~ Tabic-3-2 INSTRUMEMT LINES MATERIAL SPECIFICATIONS CC.'G!ENT a. Pipe: ASTM A 376 scamic'ss grade TP 304 stainicss stec1,1" and 1/2" schedule 80 Fit tings : ASTM A 182 grade F304 3,000 lb. W.O.C. forged stainicss steci screwed or socket welded. Sockolets: Type 316L stainicss steci ~ ~ Check Valves: Stain 1 css stcc1 type 316 with internals (spring disc) of cor.: pat- ) ibic stainicss steel (300 scrics). Manual Valve: 1,500 lb. ASA type 316 stainless steel globe needic valve. ig l-v k, t -x- -, ' ~, ~. ~ .~ ~~

I I An indirect indication of EFCV position is provided by comparison of p readings on redundant instruments installed on other lines. In addition, the fact that the EFCV is open is verified af ter each condition that could I l have produced a pressure or flow distrubance. These conditions include leaks downstream of the EFCV, venting or draining of instrument lines or instruments. and/or unisolating an instrument. In addition, each year the automatic isola-tion capability of'cach' valve is tested in accordance with Technical Specification changes issued in November 1971. Based on the operating experience with the instrument lines at Oyster j l-Creek there is reason to belicyc that they will function properly during either normal operations or an cecrgency. 3.3 Instrument Line Break Analysis Because of the high quality of materials used in the systems and the lack of vulnerability to mechanical damage, a rupture outside the primary containment ' upstream of the excess flow check valve does not appear to be likely. Neverthe-1 css _, this accident is analy:cd as follevs in ordcr to demonstrate that the i effects on the health and safety of the public are minimal even in the event of such ' an unlikely occurrence. These analyses are reported in Amendment 68, " Application for a Full Term Operating License". 3.3.1 Initial Conditions: l a. Power level 1930}Nt b. Reactor pressure 1020 psig c. Reactor Coolant Iodine at Technical Specificatica Limit 8 pc/gm d. Noble gas stack release rate at Technical Specification Limit 100,n00 pc/sec y g % -,==,- P . gy -,, yr q, d -f g y a m r -,v-y e r 7 ,.v ,9

) ' 3.3.2-Assumptions: ~ The discharge rates and the total discharge values given,in Table 3-3 are based on the following: a. Instrument line penetrates near the bottom of reactor vessel (maximum subcooling and pressure). b. Instrument line length to break - 30 feet. ~ Once the coolant is discharged to the reactor building, the following is assumed: c. No condensation in the reactor building. d. Reactor Building isolation and the standby gas ~ treatment system (SBCTS) are initiated by high radiation in th'c ventilation ducts. e. The leakrate out of the reactor building is . assumed to equal the dif ference between the instrument line discharges and the SBGTS flow of 2600 SCFM. (Sec Figure 3-2) NOTE: The minimum expected cxfiltration rate is equal to the measured infiltration rate at a P of 0.25 inches H20. This leak rate, increased linearly with pressure, was used to calculate the maximum pressure in the building. The calculated peak pressure in. the reactor building is 3.3 inches H 0. The building and associated syster.s arc designed for 1/4 psi 2 (7.5" H O) internal pressure and therefore containment integrity is maintained. 2 - The maximum laak rate is used in the analysis for conservatism. f. All the iodinc in the coolant that flashes to steam is assumed to beco=c airborne and leaves the building. s, G e

g,

4 ~ ~ - ' - ' ~ v-

s. TABLE 3-3

~

COOLANT F1.0tJ FROM INSTRtNE!T LINE BREAK 4 Coolant Flow Time Step Coolant Flev Rate During Time Step (hr) (1b/hr) (1b) 37,400 O to .5 74,400 .5 to 1.0 64,300 32,150 - 1.0' to 1.5 54,500 27,250 1.5 to 2.0 47,000 23,500 2.0 to 2.5 42,200 21,100 2.5 to 3.0 39,000 19,500 3.0 to 3.5 36,200 18,100 3.5 to 4.0 34,000 17,000 Total Discharge 196,000 + ,c,, e v -em+,- , yo y ~W'

}7<v"-

v' ~. fyK j

The thyroid and whole body doses are calculated using the following assumptions: g. The equations given in the LOCA analysis Section 3.3 of the Oyster Creek Application for a Full Term Operating License, Amendment No. 68. h. For clovated release, a stack height of 110 meters and fumigation ~ condition at 414 meters in accordance with' Reg.'Cu'ide No.'l.03. 1. For release from the Reactor Building, a Pasquill F and im/ scc wind speed with a building wake factor of 3.0 in accordance with Reg. Guide No. 1.03.

j. 50% plateout of iodines in the reactor building.

~ k. 90% SECTS filter efficiency for iodines. 1 1. Breathing rate of 3.47 x 10- m /sec. Noble gases corresponding to a stack release rate of 300,000 pCi/sec m. ~ after a 30-minute delay arc assured. The assumed flow rates of noble ~ gases after a two-minute delay are given in Table 3.4-1 of the Steam Line Break Accident section of A=cndment No. 68. l To obtain the released nobic gases into the building, an average flow rate through the break for the four-hour period is computed from Table, 3-3. The n=ount of gas assumed to be released is then cc=puted by ratioing the full steam flow two-minute delay values (Table 3.4-1 of Amendment No. 68) with the average flow through the break and integrating for the four-hour period. Noble gas isotope characteristics for dose calculaticas given in n. Tabic 3.3-2 of Amendment No. 68. ~ ~ ~ Gas release rates from each potential leak path are shown in Figure 3-2. . o. i e 4 - l _m. , -_. - _,._,__7 e _ -. ,mm

O 10 t i.' ,i ' i .i ti itjit. i i... iti i. . it

s

'1 1 I* l 'I!! i ~ i!'I iI'I I'ii F' ' i s H! i fl' 'l' ti i ! i. ! I' l ! !- })H l i.; ; ii, .i ei I -1. i [ [ i ! i ':i i.ti

!. i l'...l....'...'ij i'.

j!j! i ' lj j. :. i* !il' i d .-llJ.. t. l.. '...' ',.. '.. 1....'.. ( I

~'

6 I. 1..4 . !.-.-.... I..\\.. -....>1. ..l ........ r. .t.1 .I!.... I f...t+ 4 ,a l ..l - l. t.!....... .6.. 6. 5

_.-6.{

4 I..*.* .'*.'l. ...v, ..I ;.+4*r. I. *,..- ..-"4-p... .... I. .,.... 6 I l.. ..I........g g.!.... t.. .-* *f-t, Z . +p.+ g. p g. l..+, _ . 4.4 4

4. 6

e. ;

n i i,

'., [ a.
.: 0. I t :- ;
i...: ..

i '-. :<

d. :::-

n.r.;.. e.. i :

.,i

1, p' }
r i..m.

.r l s.. . i.. L..w... s.....l. -t. ..n s =. ..:._:=_:_ \\

.;.2.:.

. i =. :.

i.... i.s.
s....

a::. 1 1: 3 TO AL COOL l A.,JT......... ......_.._m.... _.- - r. ..i..- ....i .......i ....r i .t .., _ ~ gh. e 0 1 4 .,...t,. 2, T,. .. l.. l.... ...,. _ _.. 1._,. .t.._..t.. _,... _. t.... i _m_ ___r.. a ,. t..... ......._...I.:_.4..,- .~ . 6. a -.a_.. ~ ~. :.- i .v. a q.., - . L. {~_ _. ...i.:.,_ ... T. g,_;.. : -...i. _...s. ..t. ..._4 -l.._ .t. _t, _.}_.. . ~l t.;..p.._..:.... .; g..!4 -.i. ;.; .....h _-M, _ _ L_ _4.4._ ...,._.. p. . _L.Lu ;...H _:- _i. g. q,...{_g e .i i L t -t - .i 4_;.p_. i...:- a +,. .p._ f 1 ~ _: 1.. .i.;. ..a g 4.4.- _.g.;.,.t. _: *t _u. ;. 1.t..l. j. ...pt. -l l.*.j...Li h'.:. .u.[' ~h.. yg.; j.l_

_p.p _1'.l.-.

-Q.yL'.{.L h. ~ _..L;.._. _L ~'.t ...s. i.. .l:...._ .._ _}-_-_qt.j_ .I _pl '. .j_j_. .l.. I,l. .1 *_ l- -,_l..,_ 1. 5 lii. .l. , l _.. ~ _s.p..I _i . l..._ J..,_i. i. ' .....i.._. 10 .i ':3.i.L ..3,. . i i i - _! ~'... .i' 8 ' '7 r...,.a - *i.y - ' '. l.'

4 4J f-p*

a i.: : 's- _p : fi l 9 7 ". g i in 6 6 r.- ._0h-:.. p i-d-.4 : }9 e.p_-p. + ++ - 1 :.:. : y r i p.+g

4. i N w.:

. %.1 u-; e c 3,u slM. -::4 H.: ,4_;= l ai:.:

H : l:!:i=_-.= ::!.W l.!>::= :#-1=hbi.'

H - :-b "..i e...._.... : 4 o ........_.,-...._..4,....... t... _.. _. _.... s. o g o _.L... k.,_._. ;........ .t..._..._....._......_...t__._.__._. .. _,.. _.... _._. ~.._.._.. y a._ ..__...._...._t_.......... ~ c: o.,3 ___._..i.._....__l____.._._...__..._._._.._ .l.... ..._._..I..._._........ p. 1._.a... s. 4_... _, _... g y_o_ . _. g. _. .....,., _....{.....__,. .. a_ ..._...... i. 4 .__a. .....z. r.... : ;. u... p.:... l . -.... l..... n. r. ; _- =. \\... : = :.- r.:,:.: :.: : : l=.:_.. hl 1,..-, l ..:_v

.-~ m.-

.. _.._ _.. =_= = =- - .....,=..._..__._=.__n_ s:. __ r l c: . - _.............. _. _j ..._._3._..__...,_..._.._.....__..._ n.... _...... .....1.... s,,, y_ ._..._.t_.._.__.._ 4,. L _.__._,..._-..._....j....._ ....._,...f............__...i._...-.....____t__...... . t._.1,. : a ...t_.._..... _. _. o .i _.a 4, .w u. L._... _.... _ _ _. _,........._. _.....,... _, _a._....,. ..L__ ._e 3 u. ..___L.._... ..._q._..._t_....._.__._..=..._ m..._._._...

r..,.. e..

.._.4._....... -.'._TO T,*,L ST e.".,... o" c_ LC/o".'#.D W_ '.. _. .m._. 4.__ 0- /.. p ' _. _.a. . '..__ h _L . 4 _g. 9_. j._ .'O.. a. l. M. _. _._

4

. ;_._4 L_ . 4. 4, -t-.. -._---

n. o

__ I.. t..1-.,, .,. h. _.-.a.-- 5--"----- ,-F---.-t-*--- g ...s--- o;i s.T_ E.. A.m_ F.L. O,'P...,.T..H.RCU __.1._i-H_p-H_L. 7. -!-H 2- -H.. - W M_ _p:_r.u- _:;.t-.. -, l -H -}.i..: M_;T_ v

_l

.:..,.2_t a_Lu:...._u_y J _;. L:_. .t t,3

i. !

it i i ;: . TT.]._'.i _L Lu._ a.,..La _ q. L .i i..c s i i,...;ii..:_ _._.1.,...I. 1 ._.4 i . i _. i.....i... !. o,p., 4 l.. l _.: _ t LL ,..i _ i.--.J. i..i. i.. .,-li t i f-. it.:

i

...i i +ii t p: go 4, j ' - 4 . J -i i g - p._.- t -- d t.g3 +. i '..iSTEli.1 F! o( ; i-yi .,-+ p. m-i,

c..r, f, j h. 3.L. ;n,- F.'l TGCUG l -!.l.

4.. u. ariY S Ls...b..,,. 6 4, .. l.. ; n. _ m._.,e n. :. o g., ..;o

:n:j n,...::

t..

r

.r..-_ -y

t.

../..... p_......~~._

...:.c
r

.....r._~...-**'.'..."*.... _.. .... b.... s ....I w w 6- . 1.."*.'*....'...' u ....,._.1... .....,. /......_, ..r....,_._.. _. -.. _...,.. .,...t. ..... _i.... N 6.. S- ......._...i..... l.,.......

9..._....,.1._......__,.._..._. _.

v . L.. m. _.. /. ,.........i. 4. ....2.. ..$.i ..a.,_. - hi ' _._. :i M' ~. T OYSTNR' CREK WUCLEi.:.H-. J f'[_ u...' t - 9 : -'l ' i i t...- -.. - G c...J. AT_.,,., G,;> i_/u,. I.O n, j ,. :::. ;- l.. ..l-- 1, e .i __. i s-- .f........ N,,,,,

. I N_ST R U_W,., T..'_L.'.'.."c'* ' h R E.*. n'

1..

t .l...<... ._...4.. f,... .... ~1.. ... Q y3ig-l.. l.. ._..............4. .,.6 i - _., v4A..,.a,,-( !.. 6 4.,.a_w, L,.,_..N___ 1 . g 6 4...g

7..

l ;... , _..i.

g.,..m ig..

g ,3.,g. .j l,i.,.. ./.. ;. ..l.. ..l:

a.;.g.

,,,. J

i. a.:

.i.L.. L.. t. - ) 4.p. 6 ,1 j y 4_. /.. 4. m G U., c, ;.,_.- 3,2 --] .), ...e .. ; i.,- ..i... ......4 i . }.. s. l.. !., J,. }. 4 4. H.. . i. p..; .i : i...,.... i. l .l.7 g.t .,.i...i...i i,. ..6., ..4 f. -l.. - - + 3, 6 i ..i i. i'.- t. - t ... i, . i i / i., r l.'i L I- ': - ' t t- 'l '- l, e-.- -+-*i ii

i. i i

4 i i i. lo t i, 4 1 2 3 4 5 G 7 Tl?*E A ri '~ n P fE /il' (Er.) - ~ m. w,.., ( - - -.. ~.

3.3.3 Accident Sequence 6 The break is assumed to occur in r~.c worst instrument line, i.e., with shortest length and greatest subcooling. Full flow discharge continues for one half hour although it is clear that action would be taken much sooner sincc the leak would be detected by one or more of ~ the following: By increase in area radiation monitors in the a. reactor building, and S3GTS initiation, b. By coraparison of indications in the control room with redundant instruments on different lines, c. By high or low alarms in the control room on the instrument af fceted by the break, d. By half scram if the rupture occurs on a line serving protective system instruments, c. By the sound of the steam discharge or visually during normal hours, f. By alares in corner room floor sumps or increased collection in floor drain collector tank, g. By increase in area temperature conitor readings in the rc. etor building. Even though all of these indications are availabic and would undoubtedly result in early scran initiation, it is assumed for conservatism, that the scram does not occur for thirty minutes. A plant cooldo.m is then co=menced at the normal cooldown rate. At no tiac is the core uncovered. The discharge rate decreases over a four hour period until the vessel is depressuri:cd and action can be taken _ _ _ _ _ ~ ~ ~ - - - ~ l

r,,.

}- s to plug the Icak. During this this time all of the-fission products re-leased in the steam are mixed in the reactor building atmosphere and leave 4 .dse building through the SBCTS or leak out through various leak paths. The split of flows is shown in Figure 3-2. Even if the Icak rate is not this large, calculations ~of pressure buildup assuming no condensation and leak rates 4 equivalent to those measured during the~ secondary-containment leak rate test, shob that the pressure reaches about one half of the design internal pressure for the reactor building and associated systems. 3.3.4 Results The 'results of the ddse calculation are shown in Tabic 3-4: Case 1. Doses af ter 2 hours based on the evaluation above. Case 2. Doses for the course of the accident. __ Case 3. Same as Case.2 cxcept SBGTS is not used. ] Table 3-4 Results of Off-Site Dose Calculations for the Instrument Line Break Accident (rem) Low Caoc Site Boundarv Population Distance Numb er Description Whole Body '1h ry oid Whole Bocy Thry oid 1 2-hour release 0.0241 1.03 N/A N/A 1 (flow split per Figure 1) 2 4-hour release 0.0288 1.08 0.0015 0.042 (flow split per Figure 1) 3 4-hour release 0.0537 - 2.39 0.0019 0.087 (all leaks thrcuth b1dg.) .y v ~~ s

. s ( - In all cases the doses calculated are substantially below the 10 CFR i 100 guildelines. 'Ih crefo re the lines. are si:cd to (1) limit the discharge to within the capability.of the make-up systems, (2) to maintain secondary 1 containment integrity-, and (3) limit potential off-site exposure to well below t ' the gu'idelincs of 10 CFR 100. 4 I e, A w W ~ r d e== 0 e t 'I_ e 4 l ~ i 1 r - - - - - ~ ~ - - - .'b~ .. - +. -, .,--r c,,., e <,.., [,-

I 3.4 SUMMAFY COMPARISON BET'JEEN OYSTER CREEK INSTRUMENT LINES AND REGULATORY GUIDE 1.11 and SRP 15.6,2 Each paragraph of the Regulatory Position from the Regulatory Guide is reproduced below and the details of the Oyster Creek Installation are presented next to it for comparison. REGULATORY POSITION OYSTER CREEK STATION To implemeat General Design Criteria 55 and 56 Cencral Design Criteria 55 and 56 were not in for instrument lines penetrating or connected existence when Oyster Creek was ' designed. How - to primary reactor containment: ever, to meet the requirements of codes and good engineering practice of the day to protect the health, and safety of the public, and to meet regulatory requirements of the day: i I 1. Sensing lines for instruments ' that are part 1. The instrument lines at OCNGS that are part of the prote,c, tion system: of the protection system were constructed so that they: g Should satisfy the requirements for a. Satisfy the requirements of redundancy a. redundancy, independence, and testability independence and testability of the in-of the protection system. struments associated with' the protective system. t' g Y 4 (; O C)

i b. Should be sized or orificed to assure b. Are sized downstream of the excess that in the event of a postulated failure flow check valve (1/2" line) to mininiie of the piping or of any component (includ-the discharge should's line' rupture or ing the postulated rupture of any valve be broken. Such leaks will be automatic-body) in the line outside primary reactor ally isolated from the reacgor by the excess flow check valve. Upstream of the containment during normal reactor operation,' (1) the Icakage is reduced to the maximum EFC valve.1 inch piping is' used. While extent practical consistent with other the flow out of one of these lines, should safety requirements, (2) the rate and ex-it rupture, is not theminihumconsistent with other safety requirements). The in-tent of coolant loss are within the capa-bility of the reactor coolant makeup system, i strument line break analysis presented in (3) the integrity and functional performance paragraph 3.3 demenstrates that (1) the of secondary containment, if provided, and rate and extent of coolant loss are within associated safety systems (e.g.,

filters, the capability of the reactor coolant make-standby gas treatment sys tem) will be main-up system (2) the integrity and functional tained, and (4) the potential offsite ex-i performance of the secondary containment -

posure will be substantially below the and ass'ociated safety systems is maintained guidelines of 10 CFR 100. and (3) the potential offsite exposure is substantially below the guidelines of i 10 CFR 100. l c. Should be provided with an isolation valve Are each equipped with an excess flow c. 1 capable of automatic operation or remote ep-check valve that is automatically: shut cration from the control room or from another l when the outward flow of coolant exceeds appropriate location, and located in the line 2 GPM. All of these valves are located outside the containment as close to the con-within 34 inches of the primary contain-tainment as practical. There should be a high. ment. The simplicity of these valves and degree of assurance that this valve (1) will operational experience provide a high degree not close accidentally during normal. reactor og. assurance that they will both remain open op e ra t ion, (2) will close or be closed if the for normal operation and close when required. Instrument line integrity outside containment The valves can be easily re' opened once integrity i is lost during normal reactor operation or 1 A self-actuated excess flow check valve is acceptable as. an automatically operated valve provided it has all other features specified in the guide.

I c. (cent'd) c. (cont'd) under accident conditions, and (3) will has been restored. The status of the reopen or can be reopened under the con-valve (i.e., opened or closed) is not ditions that would prevail when valve indicated but can be verified as dis-reopening is appropriate. Power-operated cussed in 3.1. k valves should remain as-is upon loss of pover. The s tatus (opened or closed) of all such isolation valves should be indi-cated in the control room. If a remotely operabic valve is provided, sufficient inforcation should be available in the control room or other appropriate location to assure timely and proper actions by the operator. d. Should be conservatively designed up d. Arc conservatively designed up to and including the instrument (entire. system) to and including the isolation valve and i of a quality at least equivalent to the and are of a quality equivalent to the containment. These portions of the lines reactor coolant system. As discussed in should be located and protected so as to 3.1, the lines are located in areas minimize the likelihood of their being 'where likelihood of damage from falling objects or mobile equipment is minimal, damaged acciden tally. They should be ) j protected or separated to prevent failure !!alf of the instrument lines exit th e d ry-1 of one line from inducing f ailure of any well through their own penetration and are other line. Provisions should be included therefore well separated f rom each other. 4 i to permit periodic visual inservice in-The remaining lines are in groups of 2 or 5 lines. i!oweve r, the high quality and spection, particularly of those portions i of the lines outside containment up to and conservative design which make the rupture including the isolation valve. g of a line so unlikely, also give assurance that multiple ruptures will not occur as a res ul. t of one ruptured line. All sections of the lines outside the drywell are accessible at all times. l i 4 o g 20 - t s 1 ( Q 0

~. i ,I ~ e. Should not be so restricted by components c. Are not restricted by orifices or '] other components in the lines such that in the lines, such as valves and orifices, i th at the response time of the connected the response ti=c of the instruments is instrumentation vill be increased to an un-increased to an unacceptable degree. acceptable degree. 2. Sensing lines for instruments that are not 2.- Sensing lines at Oyster Creek that are not part of the Reactor Protective System: part of the protection system: a. Are identical with the lines described a. Should meet the provisions of 1.b., 1.c., 1.d., and 1.e., cbove, or above and therefore meet the same criteria described in 1.b., 1.c., 1.d., and 1.e. i ab OV e. b. Should be provided with one automatic b. Not applicable. isolation valve inside and one automatic valve outside containment. "Ihe valve cotside should be located as close to containment as practical. J l e I i a l 1, e 1 i ( t ,e l l l 4 21 1 1 i 1 i

. ~ Standard Review Plan (SRP) 15.6.2 contains additional requirements not included in Regulatory Guide 1.11., Compliance of the Oyster Creek instrument lines with these additional requirements of SRP 15.6.2 is discussed below: Comparison with Oyster Creek Requirement 1. All piping beyond the primary containment Section 3.1.2 of the Oyster Creek FDSAR classifies wall'up to and including the outboard all of the piping from the reactor vessel to and isolation valve or excess flow check valve including the first isolation valve external to L must be seismic Category I. the drywell as seismic Class I. 2. In the analysis of radiological consequences of breaks in the instrument lines upstream of the isolation valves, the following assumptions should be made, j An iodine spike occurs and all of Iodine spiking was not considered in the o the iodine in the coolant which analyses summarized above. The effect of flashes to steam becomes cirborne iodine spiking on the calculated dose rates and available for release to the is examined below, atmospi.cre. The mass of coolant released during The analyses' presented were based on choked a two-hour period is estimated with flow through the shortest instrument line l o with the maximum sub-cooling under normal the assumption of choked flow at a fluid enthalpy equal to that of the conditions. Both 2-and 4-hour periods reactor coolant under normal operating were evaluated. conditions. j i i t I l ? 22 - i I 4

=... ~ l r e i j e .s i Requirement Comparison with Oyster Creek o The initial fission product concentrations The fission product concentrations used in the in the coolant should be the maximum analysis are the current Technical Specifications equilibrium values in the Technical Spec-limits. The effect of an activity spike'is ifications. The effects of an activity discussed below, spike is modeled by increasing the activity release rate from the fuel by a factor of 500. Fission products should be assumed to be The instrument lines do not penetrate secondary o released to the environment if'the.line containment. Fission products in-the coolant i carrying the-coolant penetrates the , which flashes to steam were assumed to be secondary containment. Otherwise, the released to secondary containment. The fission products in the coolant which containment integrity of the secondary containment flashes to steam should be assumed to be was checked for this transient and is maintained. released to the secondary containment. The integrity of the secondary contain-ment should be verified. Depending on the operability of the gas Analyses were performed for a split in releases o treatment system, ground level or elevated from the building and the stack (Cases 1 and 2) (stack) releases should be assumed. and for the case where the gas traatment system was not operational and all of the fission products were released from the building (Case 3). The elevation of building releases was at ground level. The elevation'of releases from the SBCTS was the top of the stack. J 4 e j h 1 i

...t-o 3.5 Conclusions The following conclusions concerning the instrument lines penetrating the Oyster Creek containment are dravn: The lines, designed and constructed prior to the General a. Design Criteria and the Regula* tory Guides, do not completely mcet ' the provisions of the Guide; however,-the-do'ses calculated for a postulated instrument line break are vell within the requirements of General Design Criteria 55 for the maxibum equilibrium coolant activities assumed in the analysis. The effect of iodine spiking which was not considered in the above analyses has been evaluated by revising the dose calculations to reficct the maximum Iodine-131 ,. coolant activity specified in the Bb'R Standard Technical Specifi-y' ic r *,

t '~ ' g cations of,4pCigm. Specifically, the effect of iodine spiking is to increase the cal,culp'ted. thyroid dose at the site boundary by 5

~ the ratio of the Standard Technical Specification limit of 4 C/gm j Iodine-131 to the Iodine-131 equivalent used in the above calculations of approximately 1.32pC/gm, or a factor of about 3. On this besis the predicted thyroid dose at the site boundary af ter 2-hours is increased from 1.03 rem to 3.1 rem. This value is about 1% of the 10CFR100 limit and is acceptabic. b. These lines were constructed in accordance with the specifications and design criteria of the day and high quality I materials and appropriate testing were utili cd. Operating experience with the isolation valves has confirmed c. their design and operational reliability. T - -7 y

- 7.- o d. The results of the very conservative instrument line break analysis show that the core is not uncovered and the secondary containment integrity is maintained.' Present surveillance activitics are more than adequate to e. assure continued reliabic operation of the excess flow check valves. Therefore, there is reasonable assurance that the present Oyster Creek instrument lines are adequate to protect.the health and safety of the public. D m 9 .e W M M M m g m 9 M e e-9 25 - I . -_ n a

, r* i o d. The results of the very conservative instrument line break analysis show that the core is not uncovered and the secondary containment integrity is maintained. Present surveillance activitics are more than adequate to c. assure continued reliabic operation of the excess flow check valves. Therefore, there is ressonable assurance that the present Oyster Creek instrument 11nes are adequate to protect the health and ~ safety of the public. G .e.. e. O e e4> m 64m. 6 6 25 -

mi SEP.To'pic:

III-3.C Inservice Inspection of Water Ccntrol Structures . Safety Objective: "To assure that water control structures of a nuclear power facility (e.g. dams, reservoirs, conveyance facilities) are adequately inspected and maintained so as to preclude their deterioration'or failure which could result in flooding or in jeopar-di:ing the integrity of the ultimate heat sink for the facility." / Introduction e-. The Oyster Creek Nuclear. Facility receives its cooling water from Barnegat Bay via the South branch of Forked River and an intake canal. Water is discharged back to i Barnegat Bay via the discharge canal (Attachment I). The intake / discharge structure houres Trash Racks, Intake Traveling Screens, Service Water Pumps, Emergency Service Water Pumps and Circulating Water Pumps (Attachment II). j In addition to sea water cooling, the Facility receives fire protection system water from a fresh water pond / reservoir which is controlled by a small pressure-treated wood-bulkhead and concrete spillway which discharges the water to the canal l portion of Oyster Creek. j 1 I Intake and Discharge Canals 1-At least annually for both the intake and discharge. canals, a bathymetric survey is conducted to determine bed aggradation or degradation and siltation from the mouths of Forked River and Oyster Creek up to' the Facility intake / discharge structure. This survey has in the past resulted in dredging operations to restore the condition of. this water conveyance. a In addition to the surveys being performed, bank stabili:ation for portions of the canals on the west side of Route 9 has been provided. This stabilization included placing rock in the vicinity of wave action and asphaulting the slopes. The condition of the embankments is periodically inspected and repaired as required. ] Intake and Discharge Structures The intake is periodically inspected for conditions that may impose operational constraints usually during refueling outages. Divers are used to check for silt and debris accumulation in the.immediate vicinity of the intake structure. The pumps at-l the structure are inspected for accumulation of marine growth and periodica11y' cleaned. The concrete tunnel which supplies cooling water to the main condenser is likewise inspected and. periodically cleaned as required. Operations personnel on at least a shift basis inspect and cican intake Trash Racks. The Traveling Screens are inspected for proper operation shiftly. Screen DP alarms are provided in the Control Room to warn the operator of unusual build up of seaweed / grass on the traveling screens. i - - - ~ - e ~ -- -c,-

1 e 2-Fire Protection Pond and Intake ' tructure S The condition of the fire pond bulkhead and spillway is. inspected during the , monthly fire protection inspections. The intake structure is inspected daily for any build up of debris.by operations personnel and cleaned as required. The pond has been initially sounded 'in 1976 to determine the adequacy of water supply. This survey is scheduled to be repeated again this year (1981). The results of this survey will determine any corrective action necessary and the periodicy of

future surveys.

Flooding There are no water control dans in the Facility cooling systems which if ruptured -would cause flooding of the Facility. Flooding of the Facility due to wave action was addressed in the FSAR Amendment 68 Responses to AEC Questions (9e and 10) in support of Full Term License. Based on the.above, it is felt that the intent of Regulatory Guide 1.127 and the safety objective of the SEP Topic-are met. g 0

e h*

h O O e l i - +... -,,, - - - -,. -,,,,. - - - - ~ ^ ' .---r- ,.,m,,y .; s

r. -

gr_ . c m-,

!~- ATTACIDfENT I Cf848 " O I OYSTE R CREE:t (2 ) f FORl!EO R IVE R ( OP ) ^ f!ORTI:S Q L Y V.'l H D "l S U I.it. E R - S T R A T I F I E D O IlGR t.t A L 1. ~ l \\ U E ,~ FORrfD h RI L'ER l g O YST E R 1.'. i t:. G {t' % ts STAllOrt ;.rf Q gi'# i N o. GC %s 3%8 TOR:".E0 A ia RIVER NUC'_ E f.3 / / [1.N-STA TICIt j p D A n flE GAT / DE A Cit / I.5* St g D A R f4 CG AT / If4L ET L 18 G [ N

LIITL DE ACit 0

E ~ ~ ~ ' ~ ~ O I/2 1 2 it {p,6.vc.S.:-..:-4 LCt.LE !!. tile 0; e r i 't uur.w .,m

j ~ e. p a ~ 6 TRAV(LitiG 2 [NTRG[rtCY 6 TRA5H WATER SIRVICf RACKS InouG SCRE C!i5, W A f f it P utt P', \\ i F0t!R CI RC:lt AI!?;r. .j / 1

\\.;.E\\ j_f /

1-WAtta Pu"r5 l / O f q: _o/ l p g Il C.. n .rnvlCr l; y-C~}h, ~ : C o n ttR {} l ruse ] l-l O .{._ _, e-A~ Ital Akt b o SCRLl?i ~ cat 4AL ^3" EUNE ( _3- ] in. 6*' 50 CI PC. WA f r R <gO5tRv. l Suri'tr ru'sstt to 4AICR P U. T U Rilli:0 Du!!,DifiG ( ~,jC ..Q () l l O\\,. l 1 ..n 1 j .c _\\ k 65 P A 510 4 STOP D ter,5 i o t or,5 / SCREEN g t oc-I C C Cri!I T 40t WASH f.C AG t !!C Y PC C I N Cul a r i c's O! 5 Cll A Ri f 'trC wAILR i RDM CIRC. WAftA PUrtPS DISCH. T u tstiE t ( IlfrAKE STittfCTl!RC Pt.A?i AT CE!rrEltLiliF. ~ 0F REC 11tCllt.ATIO 1 Tl!!itlEL u e _ y - 4.. .. ~ ~ -,. hm

..n n. SEP Topic III-7.D ' Containment Structural Integrity Tests Safety Objective: To assure that the Licensee's design and' constructive methods provide a struc-ture which will safely perform its intended functions. Introduction The Oyster Creek Nuclear Facility incorporates a steel pressure suppression containment system with a Drywell having interconnecting vent lines to a Suppression Chamber (Torus). This system is intended to provide a leak-tight enclosure for the Nuc1 car Steam Supply System and any steam or. gases that might be released. The Drywell and Suppression Chamber were designed and constructed with comple-tion in.1965, in accordance with the rules in Sections VIII and IX of the ASME Code with Nuclear Case Interpretations 1270 N-5,1271 N,1272 N-5, and other applicable case interpretations at the time. The design pressures for the Drywell and Torus are 62 psig and 35 psig respec-tively. Current Review Criteria 1. Standard Review Plan, Section 3.8.2 2. Article NE-6000, Section III-Division 1 Structural Integrity Test The procedure and test results for the Oyster Creek Facility containment vessels are included in a report prepared by Chicago Bridge S Iron Company for Burns 6 Roe, Inc. entitled " Initial Overload Tests and Leakage Rate Determination of the Pressure Suppres-sion Vessels at the Jersey Central Power 5 Light Company Nucicar Power Plant Unit #1, Oyster Creek, New Jersey" dated March 1966. A review of this document against the requirements of Article NE-6000 of Section III-Division 1,1974 Edition of the ASME Boiler and Pressure Vessel Code was conducted and found to be in close general agreement. Some very minor differences between the code and the actual test were found. They are as follows: Code Requirement Actual Test A. Pressure shall be gradually Required test pressures were applied to not more than 1/2 (71.3 psig and 40.25 psig or of test pressure, after which 1.15 x design) reached but no it shall be increased in 1/10 documentation exists as to how of test pressure steps until it was applied. reaching the required test pressure. .~..:--

- +.R I i . Structural In'tegrity Test (cont'd) Code Requirement Actual Test L B. The range of the pressure test The ranges of the test gauges ~ I gauges shall in no case be less L were 0-100 psig for the. Dry- -than 1'1/2 or more than 4 times well and 0-60 psig for the that-of test pressure. Torus. The proper ranges for the gauges to meet the-1.1/2 g times rule would have been 107. and 60.4 psig respectively. Conclusion i Based on the above review, it has been determined that the intent of Article - 1 NE-6000 of Section III-Division 1,10CFR50 Appendix A and the Safety Objective of ' SEP Topic III-7.D have been met. i ~ i

==

^

e \\ 4 I l \\ l i i <,,c-

% Es g-SEP Topic: V-5 Reactor Coolant Pressure Boundary (RC'PB) Leakage Detection e Fafetv Objective "To determine the reliability and sensitivity of the leakage detection syste=s which monitor the reactor coolant pressure boundary." Reference Standards

10CFRSC Appendix A Criterion 30 " Quality of reactor coolant pressure boundary"
10CFR50 Appendix A Criterion 32 " Inspection of reactor coolant pressure boundary"
Standard Review Plan Section 5.2.5 " Reactor coolant pressure boundary leakage detection"

. Standard Review Plan Section 5.2.4 " Reactor coolant pressure boundary inservice inspection and testing" . Regulatory Guide 1.45 " Reactor coolant pressure boundary leakage detection systems" oSection XI ASME Boiler and Pressure Ves,sel Code " Rules for the inservice inspection of nuclear power components" SEP Topic V-5 Review Criteria The basis of review for SEP Topic V-5 is contained in Regulatory Guide 1.45, the salient points of which are: 1. At Icast three separate detection systems be installed in a Nuclear Power Plant to detect an unidentified leakage rate of 1 gallon per minute within one hour. 2. Leakage from identified sources must be isolated so that the flow rates may be monitored separately from unidentified leakage. 3. The detection system shoulo be capable of following their function during seismic events and capable of being checked in the Control Room. 4. Of the three separate systems required, two of the metho6sshould be a. Sump level and flow monitoring and b. Airborne particulate radioactivity monitoring c. And the third method may be either monitoring of flow rates from air coolers or monitoring of airborne gaseous radioactivity 5. Other detection methods such as humidity, temperature and pressure should be considered to be alarus or indirect indications of leakage to the containment

~ '. - +..

SEP Topic: V-5 Pegs 2 NRC Findings Relative to Oyster' Creek "This facility is not in compliance with Regulatory Guide 1.45. The sensitivity of the leakage detection systems is estimated to be about 3 GPM. 'The systems pro-vided for leakage detection in this facility meet the requirements of Regulatory Guide 1.45 but increased sensitivity of the systems is required to meet our requirements." , Response The Oyster Creek unidentified leakage detection system utilizes an integrated flow .mster which measures the total cmount of gallons of water pumped. Dividing the time period of concern by the number of gallons gives the leakage rate in GPM. In addition, an alarm is sounded in!the Control Room if a rate of 4 GPM is reached. This is accom-plished by the use of a timer which is set for 20 minutes, 36 seconds. The timer is actuated when the pump low level switch (LSL) is p'icked up and alarms if the high level switch (LSH) is picked up before the timer resets. This represents a volume'(LSL to LSH) of 81.6 gallous and a rate of 4 GPM. To supplement this system, the Plant has installed a level sensor and transmitter-in the Drywell sump. This sensor measures the water level increase in the sump. By the use of a diffenentiating circuit, the instantaneous leakage (GPM) is measured and' recorded in the Control Room. The sensitivity and rt7ponse rate of this system is virtually instantaneous and is limited only by the transport time required for the leak-- age flow to migrate to the sump. It should be noted that the Drywell coolers have a line which runs to the Drywell sump thus assuring that steam leakage is accounted for. A containment air monitor has been installed which is intended to monitor the Dry-well atmosphere for particulates, noble gases and halogens. This system has not yet been placed into operation because of condensation problems and therefore is not cur-rently being used for indication of unidentified leakage. Additional instrumentation available to the' operator for indication of primary system leakage are: 1. Dryuell and suppression pool pressure indicators and recorders 2. Drywell temperature recorders 3. Dryvell humidity recorders 4. Suppression pool 1cvel indicators and recorders Conclusion ( The present unidentified leakage detection system sensitivity exceeds the require-ments of Regulatory Guide 1.45. This conclusion is based upon the measurement of ins,antaneous leakage rate derived from the rate of change in sump level. Experience has shown that this system provides the operator with timely and accurate information

..,:. 3,. ~ Pcg3 3 SEP Topic: V-5 U. Conclusion. cont'd~ on the rate of change in-unidentified leakage. Although not quantitatively determined it is estimated that-the sensitivity of this leakage detection method is better than 0.5 gallons per minute within one hour. The diversity requirement of Regulatory Guide 1.45 is not yet satisfied since the containment particluate monitoring system is inoperable. s i e c 1 D e w e O m w o 4-5 O 9 .-x4 q, ~,, ,%~~ b.

I !) DESTCN BASIS EVENT EVALL'ATION

1. -

Accidents and Tr C.sients B. DBE Documentation History C. Codes and Models-u.- D. DBE Performance ,m - ~ 1.0 Croup I Events 1.1 ' Decrease in Feedwaters Temperature (Topic XV-1) s 1.2 . Increase in Feedwater Flow (Topic XV-1) 1.3 Increase in Steam Flow (Topic XV-1) [' 1.4 v#Startup of Inactive Loop (Topic XV-9) y /s / 1.5, Flow Controller Malfunction (Topic XV-9) 1.6 / Inadvertent Closure of Main Steam Line Isolation ,3 Valves (Topi XV-3) '{ f] O

g s'

2.0 Croup II Events [1 2.1 Loss of External Load (Topic XV-3) 2.2 Turbine Tr ip (Top

c XV-3) 2.3 Loss of Condenser Vacuum (Topic XV-3) 2.4 Steam Pressure Regulator Failure (Topic 'V-3) 2.5 Loss of Feedwater Flow (Topic XV-3) 2.6 Feedwater Line Break (Topic XV-6) 3.0 Group III Events 3.1 Steam Line Break Inside Centainment (Topic XV-2) 3.2 Steam Line Break Outside Contaiment (Topic XV-2) 3.3 Radiological Consequences (Topic XV-IS) 4.0 Croup IV Events 4.1 Loss of AC Power to Station Auxiliaries (Topic XV-4) 4.2 Loss of all AC Power (Topic XV-24) 5.0 Croup V Events 5.1 v Loss of Forced Coolant Flow (Topic XV-7)

5. 2

's Primary Pump Rotor Seizure (Topic XV-7) 5.3 Primary Pump Shaft Break (Topic XV-7) 6.0 Croup VI Eve;n(s 6.1 Uncontrolled Rod Assembly k'ithdr;wal at Power (Topic XV-13) 6.2 Uncontrolled Rod Assembly b'ithdrawal - Low Power Startup (Topic XV-13) 6.3 Spectrum of Rod Drop Accidents (Topix XV-13) 6.4 Radiological Consequences 7.0 Croup VII Events 7.1 Spectrum of Loss of Coolant Accidents (Topix XV-19) '99.2 Radiological Consequences of Loss of Coolant Accident 7.3 Radiological Consequences of Failure of Samil Lines (XV-16) \\

^1 t F 800 Croun VI?I Events 8.1 Radiological Consequences of Fuel Handling Accident-9.0 Croup IX Events 9.1 Inadvertent opening of 3'iR Safety / Relief Valve (Topix XV-15) 10.0 Croup X Events 10.1 Invertent operation of ECCS or CVCS malfunction that causes an increase in coolant inventory (Topic XV-14) 11.0 Croup XI Events 11.1 Fuel Loading Error (Topic XV-11) a w 5 s e e

==IEN'M4M N W- ,sgmey _, _ _ _. _ _ _ _ _ - pp m,

g D, B. DBE Documentation History L The Oyster Creek FSAR (Reference A) was issued in January 1967. The submittal was prepared by the licensee, Jersey Central Power-and Light. Company, and the vendor, General Electric. Most transients and accidents were analyzed in this report at a power l'evel of 1600 MWt. ~ During 1970, additional analyses were provided at power icvels of 1690 MWt and 1930 }Mt. These results were transmitted in References 3 and 4 to support power upratings. Starting in 1974, for cycle 5 operations, the core was reloaded with u Exxon fuel. Exxon submitted topical reports on their methods (Reference 6), comparisons to GE results (Reference 8), and their calculations for 9 7 X 7 (Reference 2) and 8 x 8 fuel (Reference 11). Analysis of the rod withdrawal event is in Reference 10. For these submittals, analyses at 1930 FMt were performed. These reports were submitted with the cycle 5 reload report (Reference 7). The rc: ult of these analyses was the MCHFR for each event, calculated with the XN-1 correlation. In 1975, the XN-2 correlation was developed, which includes the Tong factor to handle nonuniform axial power distri-bution effects. During operation, the MCPR is of more interest, since this gives the power margin, and the MCPR is more readily determined using the on line computer from the existing operating conditions. In Reference 12, MCHFR results with XN-2 were presented 'for limiting events. Technical i z___s .c wa.w._- - - ww - - - e -- ame-w _

y Specifications were altered so that an MCPR of 1.40 (as determined ~ ' from the MCHFR) was assured for any transient. This necessitated a Ib. power limit of 1820 MWt. The staff evaluation in Reference P approved , the use of an MCPR of 1.4 and these supporting analyses. '~ ~ Later in 1975, additional analyses were done '(Reference - 14) to directly calculate MCPR so that the full power rating of 1930 WJt could be-supported. Limiting events (rod withdrawal, turbine trip, w/o bypass,5pumptrip,andslowlossoffeedwaterheating)$'ereassessed. Additional test data obtained also allowed operation with a MCpR of 1.32 7 These values' provide a 95"/957. convidence that (7 x 7) or 1.34 (8 x 8). fuel rods will not experience boiling transition. The staff evaluation of this analysis is. presented in Reference 21. This i SER accepted use of the~XN-2 ' correlation to calculate MCPR, and use of l l the MCPR's_ given above as the design limits. ] These results form the reference analyses since reloads after cycle 6 (1976) were conducted under the provisions of 10 CFR 50.59. LOCA Analysis The original LOCA analysis was presented in Chapter XIII of the FSAR. This analysis considered large double-ended recirculation line breaks. Later review determined that, since Oyster Creek does not have a high pressure coolont injection system, small breaks may be more limiting. The effects of the emergency condensers, auto-relief system and core spary systems were in-cluded, as well as the control roe drive pumps and the feedwater pumps (if offsite power is available). , Wb

hW69 FM, W
N@

e "'U* @ MQtMM @8+ gkw sm

4 m

Y

l} . Starting in'1974, Ex.kon supplied; analyses to support their reload fuel. Modifications.to the'models were required, however, to conform to Appendix K. No credit is taken'for some systems such as the CRD pumps which are not safety-grade. I'n11975, docucentation.of the Exxon WREM-Based-NJp-BR ECCS Evaluation Model was submitted (Reference'17). This model was used to' determine LOCA perfor-mance for Oyster Creek. Staf f evaluation of the evaluation model is shown.in. j ' Reference 18. The model was-found to be in conformacne with the requirements of Appendix K. Calculations for Exxon fuel types were done, late in 1975 and early in 19,76, to show conformance with 10 CFR 50.46. These analyses in References 19 and 20 used the approved models. Two input parameter changes, use of less than design limit LHGR's a'nd use of 1007, of the spray coefficients for 8 X 8 fuel. -were approved by the staff in Reference 21. a l The spray coefficients assumed in the Oyster Creek analyscs are. based on the minimum spray flow provided to any assembly by spray distributed from a single 1-i spray system. However, the minimum flow per assembly may be reduced due to a j narrowling of the cone angic at the spray nozzles. This effect could result in reduced spray coefficients for BWRs if the minimum spray flow per assembly falls below the minimum value previously assumed. However, Oyster Creek has t modified their spray systems such that no single failure could preclude l operation of either spray system. b e gin.T= - E _.. ,p-i

[,', k Th2' Exxon analysis discussed above used as input results. from NSSS vendor blowdown-ar.alysis and then Exxon heatup results. Exxon later submitted a model'which included both blowdota and heatup calculations in reference 22. The NRC accepted this topical report for reference tv plant applications in reference 23. Analysis for Oyster Creek using this model was submitted in reference 24. .The results from this report are the reference analysis for LOCA for tha. Oyster. Creek plant. 6 h m n e +M"^

'.O .f. ,~

C.

Codes and Models> Analyses in the ?SAR were done with GE plant transient codes. For;the limiting events,. 2hich are reanalyzed for reloads, the Exxon plant transient- ~ 3 code PTS-BWR/ MOD 002 - (Ref erence 6) was used. The PTSBUR2 digital' computer code is used to assess transient performance of non-jet pump BWR's. The codes utilize the basic transient fluid" con-- l l 'servation equations for mass, energy and one-dimensional momentum. 'A point kinetics reactivity model is used with feedback from Doppler,-voiding and J I control rods. Axial weighting factors on power or feedback can be used. ~ i The program calculates fluid conditions, such as quality, pressure and flow as well as heat flux, power and reactivity. - Control" functions such as j I reactor scram, relief and safety valve flow, isolation valve closure, water f icvel controller and steam pressure regulator can be modeled. The code was ~ latter modified to include an isolation condenner model for the analysis of 'the lose'of feedwater transient (reference 26). As discussed in Section B, this code calculates the MCPR during the event. The transient analyses were performed with the following initial conditions and assumptions unless otherwise specified in the. individual DBE section: ~ 1930 MWt = 100%' power End of. cycle void coefficient fer heatups Beginning of cycle void coefficient for cold water addition events 25% uncertainty-factor on void coefficient f 1035 psig reactor pressure bounding scram reactivity curve l i 4 I n.-+ r s-mn--,- e-meee,,- w --,,-e,. ---ny-, e-r ,v e m

) The analyses assume an initial power level of 100* (1930MWt). The Standard Review Plan requires use of a 27. power measurement uncertainty. This devi-ation is being assessed by the NRC on a generic basis for all BWR/2's and BWR/3's, but an implementation position has not been established. As required by Appendix K, the LOCA analysis assumes an initial power of 102%. The LOCA codes are discussed in Section B and Section 7.1. 3 r')NM3 AAi AAUM AA"# 5 5 & c4 gggm pi:1c w: m rt r/ A m a w -rx,ta.s o +' Aso w en A"A WM- / -e. m-e+ M + +- 9 Nr y er g _ -e

pRAFI! 3/13/81 D. DBE Performance' 1.0 Group I Events These moderate frequency events involve either an increase in heat removal.by the secondary or an increase in core flow. svaran

1. l' Decrease in Feedwater Temperature r+c. (V-8)

A decrease in feedwater temperature can result from a failure of a feedwater heater. The analysis assumes an instantaneous loss of all feedwater heating with the temperature dropping to 135 F. The enthalpy reduction results in a power increase due to the negative void coc_fficient. An overpower trip occurs. PressOre does not increase r due to continued turbine demand until the reactor trip. A second case was also considered, a slow loss of heating resulting from a turbine trip. This event was found to be essentially the same as a turbine trip, since the power and pressure transient (due to the turbine trip) is over bef ore the colder water reaches the core (delay time for water to reach the bottom of the downecmer). These results as well as those for the instantaneous loss are given Reference.9, 11 and 15. A siow loss of heating not accompanied by a turbine trip causes a more gradual decrease in feedwater enthalpy, with the temperature dropping frc: 315"F to 100 F in 60 seconds. This sicw loss is core severe in terms of MCPR reductic.n because the peak heat flux is higher. The cocpicte less of feedwater heating results in a quich neutron flux increcs: and thos an overpcwor trip is reached M h, M .K,C.: p sooner. 0-U 6 G. M 'O D MdVA J"' Mf, C, 4 J4 3.) uss rmen s c4 Acm a o. n.. .__n__

9 ...n The sequsnce of euants 'for these events is summarized'below, r e Slow loss, with turbine trip Time (sec)'_ Event O turbine trip, tertiary heaters' lost 1 overpower reactor trip 1+ bypass valve opens Slow loss, without turbine trip i i ) Time (sec), Event O FW heaters lost i 30 APRM overpower trip 1 5 I Instantaneous Loss 4 Time (sec) Event O loss of FW heating ~ 5 colder water reaches the core 4 5.5 overpower trip Following the scram, tte plant is in a condition from which a safe i i shutdown can be achieve'. 1 Operator procedure in the event one or more heaters are lost is to reduce reactor power by control rod insertion. Failure to do so would make an overpower reactor trip occur sooner than it might otherwise. i s J w -- ~-- w v' g e,. - -., - r w ,m m-- -m~-r

= p AWAL YJr $ of -rHE LOSS of ficT4/ Aria AE?TidCr Evc^17 /$. EQusvats 4T -n. CURKC47~ M K C. CKt TEttA. m t.S s.s /34JED ou -7"X E Fo LLo u)ia 0- : bretic,As Coaxasq 72s/s A SSuns77c4.5 get spcc e4tTH BEJ PE C 7~ ~15 tHITtA L. Co N oi-nea 3 14H is c c c p rA o L C Catc ut A-nwAL popCL6] pl[ usco ~ /? HALY 50 trJ BE ~~f*IC t9CCED7+MC E CAI 7"C/Cf19 fo/2 74G* GVe~^2 7" /4CE T rfer SAFa. t m e 4 ee 'b er s O O I l r - ee= - -A- &, ;.. % ~ ' .e- .n

e \\ ~. ) ,/- p g'{ g., . "/ n H t.s - - ~ < (, x V. SPo e ' We Q t.1 h.-

- l N

N g g ts ,y d "'*5 W* t* , *+ g,- s .l .r L,a Es. u -(, .... M. d,.. . -.. t' 5 L..-.-. --...r A-1 =' p> g s g u

/.

r. =

c. i.

n >.1 t '.a n H Li *.)

t.4 L.,

' * 'l s >: c; i-eg ft. g : I g .; _a li

f"

.t' b. ... n ~-- B / ("" i C

f. I 6

t b

y v M C Li a1 ^- -,y~ y -~~ 1 (C k i.eL'e.s. u 3 , (t) a1. L. O H te) 'N (T. _ .* < < ro v w. t. g- ,y ,g tal L -.... -.. J .,0.._ y .c.. to , il ). - i s f / 14. l LA-lt 7 .t 6 Lt. u O M n vi O J c. w 3O u r. 4 w. r* 6 co iP g e, et ~ Ya ^ tt h tv 4 ~ Cs O Cs o t v w c> a o o o I s e

- 4 (fi eri

e*

N o l .-] 03.1UUe 30.-lN3')D3d C' .4..* 9 e .e w ,"m-m +- -

j e f .,4 o O ,s,-- - - .? l f' .+ )\\ , g 1. e _. ~ y e g . -,g v,, u 6-n M V, -~ }. N e -.i n m n y n a ~ y, P . aj w e- {} _-. h>., ca \\ I Ti r, e, a a s _d

t.

H 3 -e tt y , >: = y ' a '.,,., s >.u a a J. o. g c. .i.,d,'4,. g,t _ v t l u g tC 4 ta 'o u..D..t.J, u... i o ,/e s x = u. > t. ut a F C4 ./.. C ,w to o .:a d v 2 ] (jj tj i. _ ? -_.g._...g ' x _. g l

e. _

u. 1: w4 tc ts. m ,,[/ F o ik .A./ -l 3 e).p#, # t y .. ~==_ .%; r .2 ij = Ql 1 a r p v> e N 4 N + eu e d N

t r

+ i o ~ o e, o c, a w a o o o o ~ ~ w v, e u o ,,) 03.1U0 c.!D.lN33U3d h

<.g.

~ " " - w-M.- .,p y. M 6,, We

b-1.2 Increase in Feedwater Flow ' '.-~ I ' C <V-') An increase in feedwater flow can result from a malfunction of the ~ feedwater controller. Power increases in response.to void reduction reactivity. The analyzed events are initiated from approximately 53% power, 42% flow, since these conditions permit the largest increase in flow, and thus feed /stcaa mismatch Th,c' transient is terminated by the high water level turbine trip in 8 seconds. Scram occurs following closure of the turbine stop valves. The bypass valve automatically opens due to the increasing reactor pressure. This event has been assessed by Erxon in-References 9 and.11. The generic reload topical report for GE BWR's considered the feed-water controller failure to maximum demand as a potentially limiting transient. The event is assumed to be initiated from full power, and ~ proceeds until the high level turbine trip produces a reactor scram. The event can be thermally limiting since the turbine trip occurs from [ an elevated power level. The bypass is assumed to function normally. The increase in feed flow from full power has not been analyzed for s.me) i l Oyster Creek. However, based on comparisons f rom other B'.iR's,J results of this analysis are not expected to require changes to a setpoint or the initial MCPR. W-8., _ ____ gW a -n-.- v.-, - -. -, -..,. - - -. ~.-em.

'O' b O O EXCESSFEE0WATERFLOR(From1025;Nt) TypeVB(8x8)QxxonNucleayFuel l- + FIGURE 3.4i.2 - a OYSTER CREEK s* 4..- e. .i1. pct:EP. LEVEL i j 140 - i i 2. HEAT FLUX FECIPCULATION FLO'.4 8 i* i3 i4 VESTEL STEAM FLOW t. d 15. FEEtt.crtTEP FLOtt I. . 4 ~~ - .~ ~~i.~ .~i.. .-i.....--.-. - i..-

i...-......

120-i.. 5 S 5 1 5 S S S S S i .i i. i l 1e.o..... g i ...........4 ....p.................g.......... ~ ....s..... 80 e. 8 un w o 5 + u ~ C u ......-g...- .~...--g- -..--.3- .. - 7 LL C) - Sg 't .;i 4 n I ,. t 7 j. O., ~ 4 e u *% ? m* m. .~. a .c -.,-s~ .,re 3 u

==w 1 ( \\. 9 3 3 3 ' ~ ".r " 3 3

* - ' h 4,

40 O .t A \\ 4 4 + 9 =# + i ... q... ..%.. 2 .y.. + s ? ..g.. an ~ i 9 s 1 1 1 1 0 2 4 6. 8 10 12 24 26 10 20. D -T!?'E. !EC

O O .O L 7s............. OiSTERCREEKi EXCESSF'EEDWATERFLOW(From102bPdt).... Tyd..e VB (8x8) xxon Nucleaf Fuel I FIGURE 3.4 2 .b... ~.4 ..--.....~........----t. i 1. VESSEL PRESStFE CHAttCE (PSD i VESSEL L!ATER LEVEL (IN..) ! 2. + + P 2 t 5 0..:.... ...m.. 2 I 6 7 2 2 j i e 4 - 2 I I - -- l I1-25 - i-t + i + [ i j .r 4 -0 ' 1 ' i 1

i 4 i 4

.I. s e en .n. i i-4 r i: . 25.: - --r- - - + ~- t-i e . 50.. 1 e i l 4 ...t. .75 .b s v. i

-100 t

2 '4 6 3 10. 12. 14 16. 1 20. ' O. I p..tc.. *. ec, ~

P,. ,M a ..-a.&..... ..-o ,a,.d..... g..........up................. e N 1 O i i 3 g + 1 ? e.: ? ? j. 0,,'9*: 1 M H .........o. * *..................... 40 : e. Si r" : L) : ? 3: E1 C.: O- '.O 34 ? 6 .. M.: A 4 44J :... A: . CO. x: CO.:

==" ;

00 : 7.: 9 ..gj 9.. g.. .4....... ...... ~. ..n-.. g 4 CL? H h: >-- g >-- :- c .?. (C. .~ ty O OH ^; +

or

? 4J : s2" ~ ? c, %: 2b ..e...

. as g

...,.. g g..

I ir 6

h Q N: >\\' 1 C: 0 e--- ? CC e4 @; fi. N<

s..

W* e. U: ,4 N y' e a

u..

e,. Mi H W:

'C i 4

c. N H: p d' ; k o. M ....4.., ....4... . w.. W: W: Mi Mi W. Ui X! I W: N e, .....y....... ? i.. 9 1 ? Mj ? kJ: W* g: Ui ? s M! N WI U + Hi .4.... .B.'.. C) :. ~ u:: I tN, s j v a gy. 6 g ...,3..... V. ;: j. .p Mi -os I i .3 : a .j* f.3e. 6 m : LA. I o....... ..........g.. .p...... t.. ..g.. G3GO 0I1U3 XG 1.1 DGI '.U~aI11;O 'Jt4IllIII ww.y a e n -m

. o 1.3: . Increase'in' Steam' Flow . 'M. XV-t) 4 An increase in steam flow'can occur due to a pressure regulator mal-function. The increase heat removal causes reactor cooldown, pressure. ~ decrease and. void fo m.ation. Ne'.stron flux decreases via'the void re-activity, feedback. E Transients were considered from three-power levels, 1860 }Mt, 960 We and hot standby. 'in each case, the turbine control valves are opened ~ t to 110% of full rated. A mechanical stop prevents further_ valve opening. The hot ' standby event has the fastest depressuri::ation rate, as would be expected. The decrease in pressure nould result in MSlV closure (825 psig). A I reactor scram is initiated by low reactor water level. The MSIV ~ closure would also trip the reactor upon their reaching 10% closed.' Pressure regualtor malfunctions are considered in Secti: r. Vll-8 of the i FSAR (Reference 1) and Amendment 14-(Reference 2). Since there are only minor consequences compared to other possibic transients, no re-analysis was performed for later cycles. t 4 i l-r I S 1 L a t N_ m r

. e. WC ,4WAtyJ't,,$ 0 $~ -17/E 3RCTUP Of /dICLC /isC('C. low & :::a T--- /$. EQ UWA LO T -G CURRZ41~ MKC CKt TEt/A, 7~2 t3 t.s - /39JED oa -7H C FoLLou)ra ce- : b//itLAA. Co-J:K!2srt17Ws A SCusm*1704J Ale ~ AA0e~ t JIT'I /tc~a A E C 7' ~ $ - / Hi TIA L Co N Oi 73e4 3 ] 14H is c.c cp TA 6 LC CALC ut A -rtoMAL. popCL.(3J i+4C uscQ t& ~774E 14NA Ly' t) 3 6~ f4Ccgpi+NCE CA'lTEY2t/? fot2-7" M GVe~^J 7" f G -W e? SA M d. 4 f 4 m ( i l I i l - - - ~ ~. -,, ~ ~

1.4 Startup of an Idle Loop f,-r_.rg, xv 9) The startup of an idle loop could result in a cold water addition to the core, which could cause a power increase through void collapse. Oyster Creek is a non-jet pump plant with five recirculation loops, Procc'ures and interlocks on pump each with isolation valves. d starting and valve opening effectively prevent such a transient. Multiple failures are required for a severe idle loop startup event Co occur. Procedures require that the loop discharge valve be closed and the suction valve open before the pump is started (a small bypass line prevents dead-heading the pump). After the pump is running (at 3 07.), the operator can open the discharge valve. When it is fully open, he closes the bypass valve and then the pump can be placed in automatic - speed control. Normally a loop will not be isolated when its pump is shut down so that reverse flow via the bypass valve will keep the loop hot. A 50 F differential requires reactor shutdown to restart the pump. However,.an analysis of this event has been performed by Exxon at '1930 MWt in reference 25. The following assu=ptions are made: 1) Water in the isolated loop is 100 f. 2) The suction and bypass valves are opened at the same time. 3) Coincident with valve opening, the pump is started and quickly brought up to speed. I

s 4) The discharge valve is opened as soon as the pump is isolated. 5) 100% power and 100% flow from 4 running pumps. ~ 6) Scram setting is at 1167. power. Cases were also assessed from reduced power and flow conditions. The full power case was the most severe. The results show that a high flux scram occurs 7 seconds after the cold water reaches core. The MCPR limit is not violated. ~77/E /fCPR-F (L W'S Tv9 '# 5 ' E 'J T /5 /. 4 /, i e 4 e o w V9 ~" ,-Ww

,w j, W Xil-75-51 JI Ltya r, 2 3X3 O C. I:LC L ': P STF9;'c7 ra23193::.-u - 1* C [ ee.cr un. ---- - - ~ a + t-e r. r. t g 1 oc::e.:.arv.s r. - l - - - *L-- -t rec:.uce rce.

e. + ime; srce rte.

I 120-

- - - ~ M "'sp.g ~~}., ~s; ' s s

2..

, C. t ? 1 * * *,~,,,.s t 2. 3. d G,. L I y 3*. -2,,,. 3 2 i 3,* g.,. 1,Es2 3 -w l g s, ~ e, w m I }... m3;..... i 3 2-k. l (' e2 I -+ 'i . _...... '. - \\ 7 t 1 w3a J LJ l , \\ )

w I

I a. t .\\ .t, .gg........ -.j...... i l \\.' i. g l t l 2C - t 1 I. ,s 6 1 i l - g'- A.- 8 .. g.._.......... k.._ a

~

1 t 3 3 2 't 6 8 _. L. 12 14 16 8 23 ] 1C _,,a TIr'.C. SCO r. a ~ 8.:2-'. tr 815 O C.

'.C Lc::P ST5;T*r FA:M 193C N J

6 " a -- - { 1. veset recr..T :-e se :rs: s a ecssc. w ca sc<c. 4 + a SC * .i 2 __ _ _ 2 _2 .2 2..- 2 2.- 2N~ -2 23--~~-------- ~. } r I I I. .. I.. . I....

- - - - k.

,g s[\\. 2 C--- . c. A i - N I z 37 + .g i I ~ .sj; l 1 41 r s t l l 3. 2 4 6 8. IC 12 '9 16 18 20 s a TI.wC, SCC y.] a v. c-, n..

g O' 23 ~XN-75-51 t 1 318 G C.1]LE L32P 51 ARTuo rROM 1630 W J. ,ig pp FU'U' i 19 G s' .o

re.1 l

3 or m aatimFtm 9 etin $f f' F1% g a tt 3s t CW f t #4 y l I sl 2 g s s _i _ 100 J s, _.m_ LM', ! > 3' = 1 is i g 3 4, WSO D 9 E M60 5s 't 40 I, \\ e \\ x -x 20 1 N !N A, M ;._' 3_k.-- - ut. 1 h$ ,1 c s a 12 16 20 23 2a 32 36 50 G 11NC SCC ata O L. I LC L30P STRAltp FR3M 1683 Na ,,,,g f~ ~u ~ 'a ' "# "" ' ~ T " _ _,,,, ~ so ecsgt w its t stt {- 90 9(a ~ 20 4 l c3 1 y s t ,x / v h (*1 Ll C 1 \\ t \\. 12 1. 22 2, 24 22 3.

c

-y TINC, SCC j ...A =~-s--- -~

L. -2 5-Xll-75-51 g 818 3 C. I1LE LOOP ST8tRTUP FR3M 1735 MW Uger J_I. t,

t.. _,,,.. _

1,c 3 .o1 rtus 1 sc N arte FLft. 9. FC5T., l'c4* tt s. t a esneca r6m 12q-s 5, s t C I* =! I 30 M tea [- !as Ii I i .i, ,,s,,tt ,s- - \\..) ^ $60 I t e.g ..t... _ 1 .5.. i \\ i ^ ~- ~* ~~ %D-' s 'i s 20 l \\ '.N. ;.... -'i'! 2 s ,6 3 N...)_.. _ a ,a ~-qt.' 0 0' 4 8 12 16 20 21 28 '~Y 36 93 3 11NC, SCC 5 ur as u2 ara DC 22LC L33P STRRiUP FR33 3235 MW -l ~.. .t-nsy.. mgs.T mme rms = 60 a ecssn.evce trett 90 7, - - 20 l l N '] 's t i 1 - -0 >e- ,3 ->o \\ l ? '2 -10 l- .t 'i -.60 1 6 L .8 i 4 8 12 16 20 29 28 32 36 40 'S -M TIME,5f2 41

o 1.5 Flow Controller "alfunction Causine ' Increase in BL*R Core Flow Rate /.*' - '. V - * ** A malfunction of the speed controller can cause the scoop tube. positioner t'o move at its maximum speed in the direction of increasing pump flow, l a 10%/sec increase. A =alfunction of the master flow controller is less severe since it has its own rate limit which is less than that of' indDeidual scoop positioner. The transient is initiated from 52% flow, 63.5% power, to allow the maximum possible increase in flow. Neucron flux increases due to the void coefficient. Heat flux does not increase above the steady-state value due to thermal lag. Power and flow re-establish themselves at a new equilibrium value. The consequences of this event are relatively mild, and resemble normal load followed by flow control. No thermal limits are approached. This event was assessed in the FSAR, as well as in References 9 and 11 by Exxon. e e I I I s-r-- 3 - --- - up5 i we, -gs .n e -w g 4 g

WC AHAL y'Jr.$ o F~ -77/E FC Cd coa TCOLLft-i fR/~AL7id & VCd T /3 EQusvA LL JT CORT *AA T M K C._ C Kt TEK/A. m s.7 a f,3'4 3ED CA ~77/ C fcLLoulld(r-: bt,nicAit. Coa,:,cn /A*ns!& A SSWr/1104.5 A le~ AACC ed/TH Be~3 PE C T 46 f a: TIA G Co N o t ~n e J 5 /]H t9 CC C^P TA o LC Cs4 L C ut A '71 MAL fwpGLOl A4[ U,.f C O ~ t&

- TD E ANALY t)

~~ IRE" ricccp*ANCG CKi Ta A't A An. -7ust Gve^J7' Ac2 -rn e' .SA s d. O M 9 ., - - - ~ ~ --n n..

a t t C,. ......&..................4. ........4.. ,,,..,..g.....,........4.............. ..n..l.. ....,.3.. i s .e e i i-i. i I - l i e f g t. r 9 'j I M j r-

s i

- 4.no. - 'A 1 o 6.... .....b......... ......f...................to a u. .. 4... u .u...

u. ou M

t

b j

i

q e ;4J 4J >-=

p ? a U i

3 O

LC 7, e m. l t NC h l d O.g . y,' o.. ........... -.....,... ~ q. .M EW + 0: ? E. ^ a + 'V %M t + M e 4

O CD O

C ?" l j. ^ <J : Em

D

.. CJ M.,.- .4..... ........l. .4...... .......... -............S.. 4 a c: .O 3t;O. i i a i -I 2 Oh

t-y

-b

c. 5

.m.l a J wa C.'. w '.U .L M 23 : O '* J ( L: ; 3 H 416 S..h j

t QA >- hJ

.."*g p...-X C e u .a*. o f." = e' q p} ~* *.. a + a.a a-a + g? taJ O,*J M uJ L. .J.J. * >= J-kr* M Q* (t L e S* ,; UJ >= H t: e s. ? ..: t.E v.* 9 W s C v bJ LJ L.J J j

-*: Q I u.* La l

): I O j U_. ::..... g e4 N M T M 9 O D;. _3 1 N.i i k + 2! t_. y 4 Cfj

/

w: ? . 8i. M 91 J g; 6

s g:

y w* 'eU I i .3......... ..h. .. i..,.- eO! ui g 7%* cM a. M Ai

.e-

? i. \\ 4

g

.....g.'..%.. (..... ...g.. ...............y.. ... t..... ..t... 1 ? g V i w: gg-Ni 9 l U: 4 1 M M. 1 ,{ t s t. e ..... g L g .. vp g d o.- 1 4 } \\ 4 t % -, v + 4 t* ..*Q .....i.. ...i........ \\, ... n. v.: 3 ( i t.t gf i " T t i e i. s) m i ~ w ! ~ ....,~ ~.. - - --- . o ..t... ..i-i M t CJ O

^

O M C M m F4 ao e n w o 03tVM JO Di3h]J rj ~"- .me o. eo o.m.o.e+w.= -. m m. oe .me. __mm,...- c

g I

( l I 4 e e e e -.o. i 3. ) i i .........YSTER CREEK !...... FL.OW COUTROLLER pat. U91CTION (Miximum Flod 0cmand f re..m ?.10 ...o.o+ b O. 7g.,! FIGURh,.4 !1 o ......Typ.. VB - (8x. B J xon i ...... +..... ..u... e .a VESSEL PPE55UPE CHANCd (PSD Nuc1: ear puej ~- - i ..w 5 i i. i j 2. VEs5EL W1TER LEVEL (Irj.) t. 9 4 .I. ........u....; ..j.... 50. i t + i . 2 I I 2 2 2 2 2

2 2

2 2- + I t 3 j 1 1 - 1 .1 1 3 4 , 17 4 ] 1 1 6

e. --

j 8.-- g -0 e. a t 4 e m i 8 e i i i i 4 ..u.:o onoo..o+ ono s-o.- g-- .o.o o 9.oo ...u.- ...o .... go a o, ....o...... e. ....o n t.o.o. u2$., ? l 4 . -f- - f r I i j j 4 i j ~ .j ? - ~ - 4-6 =4-75 -

e--

6 o- - ~100 0 4 3 12. 16. 20. 24. 23 32. 36. 40. 1INE, TEC

~ O J< g O FLO'J CONTROLLER MALFUSCTION (Maxir.ud Flow Demand From l0I . i j

." Type"VB"~(8x8)" Exxon'"thelearTueTi-7 9..

.. 9. ...... OY5.TER CREEK IFIGURE3.4.!1..c ....p..... - -.., ji. 6PPROX. H-L CCRRELATION j

i.

j2.

XN-1 CCMELATION ^ I t.--.+...-~~~~g-* ? .........g..... ? j .r.... -..9... ..g._ ....g,........ 9.... a = d I g. g g ?. 4 g A I W g 4 ..... ~... -....~7.= ~. ..k.. ? s. - -................. = w ? O + ? 1 H C-W j d( 4 J ................-4.......-........l..--....-.....j.......~..p L 6 -2.... g .........-....................4.......~... g H 4. e g,3 , 2 ut e s oa 8 d(r 1 F H 2 . 2 - u

*---------e.-.-

-~~----*..~..--.~..-..---.e-LJ 4 2

+--:-

=2 9 3 L 2 -- 2 -: z bJ n en n g .? ? ...?. . i

j 2. :....

1 4 ' i - i 6 I ? l ~ + g i i .. k ...4... ? ...a. 4 ....4.... ......A.. 1 g ? ? 1 e. i. I !,I 0-4 6 12. 10 20. 24 23. 32 36. 41 TIME, SEC

o 1.6 Main Secan 1.ine Isolation valve closure (m A - < v-3) Inadvertent closure of the main secam line isolation valves can result in vessel overpressurization and loss of the steam removal path through the steam line to the turbinc. Alternate paths are needed for heat removal. The relief valves located upstream of the MSIV's open automatically en high pressure after 4 seconds to relieve the pressure transient. Re-actor scram occurs on 10% closure of the isolation valves at 0.3 seconds into the event. For long-tenn cooling the relief valves and isolation condensers are adequate for removing decay heat. For a conservative analysis of the pressure, the fastest closure time (3 seconds) is assumed, and no credit is taken for heat removal by the isolation condensers. The safety valve lift setpoint is not reached. /, pcMt 7 Rus o ni. of sit o p u o. is emc 4AWC, w:LL o c c o-J Wd consr 1a re ry p t. vc This event was analy:cd by Exxon in references 9 and 11.j The results um,g showed th't the pressure consequences were less severe than those pre-a dicted by the turbine trip without bypass event (The relief valve sizing transient). BWR's covered by the GE Generic Reload Methodology assess MSIV closure with reactor trip due to the high flux scram as the limiting overpressure event. This sequence of events generally results in a higher pressure than a turbine trip or load rejection with bypass failure (with scram on closure of turbinestop valves or control valves). Oyster Creek l l 1

a analyzes an even more severe event to assess the adequacy of safety valve sizing. The turbine stop valve closes, no scram is assumed to occur, and the relief valves, bypass valves and isolation valves on the isolation condencers are all assumed to fail. The safety valves 1 are abic'to maintain pressure below 1375 psig (110% of reactor vessel design pressure). /9 FC A at Fr?CJ v C i 47 Tdc /ca 56 L ooWo n QY $ g4O l ) (, ) gy8j IS AffAD'glMA~~([y l ) [ }, $l5 bd% o i= pe cwc vt em a s cue e eipoa w, b e e 4 ggae e, _w. g

o. o .e 4 cjs/W C AnfAl y/3 Of -f?/C.fA' ttl s7EAM Li'l[ lJ%479) YALV5 gvCM T /3 S Q Ul V A L. c n T CURKZdT AlK C_, CdiM/A. md t.5 24JED OA TdC 54LLOUJ/*J&~:' bb*to(, Att. Cods 0C/A*T?di A SCUM / ~D 04 5 8 de~ AACC L4/Toi /:c~3 A E C 7~ / && 7'IA L Ce N O / % o*J $ j i t& ME ANALV30 s ' /2. 1~M G Va~^J 7" /4CE - rM G SAMd. QE~ g4 <~c g D *vt N C E C AI7"o~/2//? cs e 4 h n 6 m 6 A W ~ f i 2 i 1 w~ .u, n

7

FIGURE

.1.3 - a OYSTER CREEX CLOSURE OF MAIN STEAM LINE ISOLATION VALVES:Type YB;(8x8) Exxtni uc car Fue.) .H ..-~.. g. .....q.. a .. 4... j.. 1,,.y: 6 i 1. FU.ER t.EVEL i 2. 4.1T F h: i ! 3. TI'I'O. f.ATICH FLOW .r_ gT.TM. u.FLC11 i. c v.s c e :....

c.. c... 3 i

l s - ---- - - s~ + - ..~v.~---. ....-.-s-- 153-e-~- +- ? ? + 2 5 +~~~b---~~--*-- +- ~-- ~ -~~-----~-4~- 125-- 5 5 5 5 5 5 5 ? 5 3 i ? 9 igj. p. 4 2..

, 7 '>,: *

ss

v,.

E e O-t. s c t f La ? 1 ,a. s \\. 4.. U ( t . t,. 75 7. \\ i .s t, .\\ N, s )' 'g e 1 ) el ~ \\. 4 4 t g g 4 4

%.M"'

I 1 +- i- -- ~~ ~ { g 50 \\ ' N \\ N, . N %;. I NM b. i ) i N i -} - {- 25 M r \\ 2

~

i i. a w 2 ( 5-k ) 1 .1 g 3 t 4 3 .w b a W'_ 6 'O 17 1: 16 11

20.,

0 0 0 T It:i-

Ce*

o o Type VB;(8x8) Exxin Nqu

FIGURE 3.1.3 - b OYSTER CREEK -

CLOSUREi 0F MAIN STEAM LINE ISOLATION val.VES t u.----.-

.---------4.----

-.--a.--.- ------u..- VE5SEL PPE55UCE CHRCE (PSI) i,... 24a ---u.------------- . 1. + i 2. \\E55EL WJEP LEVEL (IN.) s i-1 g. - - e- - t---- r - -- r-200 > - - - ---t---- i t t + g ......-...........u 3 160.......... i. 4 i i --.4.-----$.-----.- j 120. 4. j 1, .g

ro -

os -

i e -.t"*-+*+-*--*aat.-* ""+ f.- g....... i e j gQ.... 4 v .............).. 4 .. ~.. .. p.. 40 t 2 t t \\ d, 22, \\ ?. t ?2- -O .s 1 0' 9 40 10 12 14 16 19 20 O. 2. 4 6 3 e4 e. e u..- +. - -

= O 'E. O 0 Ed f. .a ...s..............&..... e -.e==36..*....... ...4............

iT ;

6 F 6* b: t, ef PD U 3<. 9..... ..........g.... .y........... ? r C" : O :, X. M ?. W '. t g t ? a M: [ g0 + 1 g-1 .b CO ; g w: CO ;. 3P. W O. >4 : 1 i, Y. s ... v .@....?... C#E. .......3.........-.......*&... W. g p A-g J-J g s 3. C L EC. LU >>. sk ** =! . u. o 4 = Z* C N. a us p 't Q l p JJ .3... e4 g... N 6 1a.3 .................g.. dC...t.. I e t.. J-u o O v) 7u ? .o M. G* e4 Q. I E,d '..

1 -

p N r ,,2 %,s. ..e =4 N .y. - O' %aJ L. = y) t. . x.< *- ? 4 e e C., Lt. o....... ? a %i 4 D A N o. ~ J u t yg.,s .......^... ..i' .... ~-. ~ e sc ' LAJ ? L&J Cf ? V.. 1 g WD ? p. v, 3 .....i.............. y...-.. O. U: es.e i n 6... .4... i . g' .+...&-. ...d... O.. vo

1

? + OM .g ? Ld n* ] D h C3 je. c . *..^ *..... .. - +.. +. * .h .A.. y.. t

- t.'

( M.M.'id ) DIWJ xm3 10.111 N3II:4*] tr.:!Initt = __ 4 y . -. ~ ~, y

O r O O l FIGURE 3.1,i4 - a OY'5TER CREEK ' SAFETY VA(VE SIZING T'RANSIENT TypiVB(8x8)ExxonNuc1har' Fuel i - - K

-t."--~~*'~~~~t."~~~~~~~-~b.'

~ ~. ~~~ ?. 3...* -i-1053-~ ,~ f. - '

==

c.,.T. fc.0.. t E*L'*L 8

? 2. -ih* P.c ; e.: .-e ULM *iA FLO*J

I t *F C.. c.,:,3.

i 3., z eL.)u lt.

~.

e.. c.:.s. y.v. :.n p.m.a l

j .,...;..........-......,................i... p o. 3..

7...

g 6* 3 l 8 d .......j......... 1. .g j.. 7$y.. .,.9..... i i !e 1 i. j;. l ,I I. I). i i. .......g.............;

I

....,..................a. ............3... a t l e .o ..t..l... ..y.. ;... l Q.)...

e N

t

to i t e

= f.I i 1 C i. J g,. j f }* 'b- { W. ..t g e. t. ( !! t.:.. ...,...i....... 4 <s3. i v ~ .. } x. l e ] .e s. r ~ t l 6 l l 9 . \\ = a t ,3 g o .. d 31,53..:... .g E 5 g I 4 j

f: \\

3 I \\. ? t 0 e t l } t 4

9 to

,a ~- f 2 l ~- - W~ -- , 9 i ..../.. ..,.....[..N-.- c. R - E q q .a r.,J..,. J e

r _~ w..~.n..

..?- .w_ - q y - 5 l r. N, 4 g .: 0 5 40 f.:0 - ? ;.) 3 c. 9

C 2 ;.-

r-

i e i I i e, t L , FIGURE 3..1.4 - b 'OY$TER CREEK j SAFETY val.VE SIZING TRANSfENT - i Type VB (858) Exxon Nui.. lear Fuel i-.......i.. ....4..... i ...4.. m.. 4............ . 3.. .i1. VESSEL. PRE 55LGE CHA!!GE (P5D !. 2. VE5fEL WiER LEVEL (Itt.) t t 240. i e [ j 4 .............I..... ........j... . 3.. ...........j... gOQ.. i 9 I ,i. t j e I i ................g.. . 4.. .......4.. g 20..a.. g 1 e' pu t t s-t 4 + - - ~ - -.. e - t-r- ~ - ~ - ~- - - ~ T-l C0. ? ? ? ? s ? 2 ? 7 i 1 .g f .t ? s .4....... ..4.. ......... 3 A l co.. ? 3 ? ? ? t O c to

1. *O 2.10

? :D 4.50 E iO 6.40 .'.20 3 - 1 lt") - ~-r. : y

i 2.0 GROUP II RESULTS The moderate f requency events in group II are characterized by a decrease in heat removal by the secondary system. 2.1 Loss of Lond (Topic-XV-3) Upon loss of electrical load, the turbine generator overspeeds, causing a rapid closure of the ti:rbine control valves. Reactor trip occurs on sensing the load rejection (acceleration relay on the burbine control valve system). The bypass automatically accepts the steam load when the turbine is isoiated. The loss of load transient behaves similarly to a turbine trip event. However, since the steam flow to the turbine is initially reduced by action of the throttle valve rather than sudden closure of the stop valve, the pressure and power transients are milder. The isolation condenser is not actuated since the pressure increase does not last as long as the 3 seconds actuation delay. Thus, this event is bounded by the turbine trip, and is not analyzed separately. 2.2 Turbine Trip (Topic XV-3) A turbine trip can result from a variety of malfunctions of the turbine generator such as overspeed, or f rom inadvertent closure of the turbine stop valves. The sudden loss of the reactor heat sink Icads to ptes-surization and heatup of the reactor. +

p. Protection is afforeded by a-reactor scram upon 10". closure of the turbine stop valves. The normal path of heat removal would be through Use of this~ method requires ~ th bypass valve to the main condenser. offsite power to cool the condenser. For a conservative calculation, removal means are bypass failure is assumed, and alternative heat These sytems used, such as the relief valves and isolation condenser. I do not require AC power. Exxon analyzes three turbine trip events (references 9 and 11). The turbine trip from full power with scram on closure of the turbine P P sA a 1 3 Q$ ta t s Pressure rn nc.r well stop valves and operation of the bypass valves.3 g i'> below the lowest safety valve lift set point. The turbine trip frem full power with failure of the bypass valves 4 .(relief valve sizing transient). Scram occurs on closure of the turbine stop valves. The relief valves are adequate to keep pressure pj a 3 p r t.\\. -Th SW below the lowest safety valve lift set point e.

g d fg'ilore ok hdt "IO'I Y f~

W g $ O.O*1, The turbine trip from full power with bypass, isolation condensers 4 i is used to size the safety valves. No re-and relief valvc3 O i r2 / actor scram is assumed to occur. The safety valves are adequate to keep a ..$. /' pressure below 1107 of the design limit. [1 The sequence of events is summarized belev for a turbine trip, bypass failure, reactor. scram or 10*: turbine stop valvo closure: i 1 & pc2h .Ct h N ky'fy G e% 0 / Y b pmArec 2 vbsd nut A hq75m)uAs p)hg.

7ntP N: T40')T O'(d AD li picci. 567tif C fia # 10 0 WP.

f, ~ - - ~ -.

l e j Time (sec) Event O turbine trip 0+ reactor trip ~ 1.5 relief valve opens f.0 relief valve closes I For the safety valve sizing event, the sequence is as follows: Time (sec) Event O turbine trip 0.5 neutron flux' peak 2h heat flux peak 2.0 safety valves have opened and reduced pressure The Exxon analysis showed that the turbine trip with bypass failure was the worst overpressurization event challenging the relief valves. The relief valve capacity is sufficient to prevent the safety valves from lifting even if the bypass is neglected. As discussed previously the pressure rise due to a turbine trip ~ without scram or relief valves can be accommodated by the safety valves. Singic failure of pressure relieving devices are considered in the analysis. Operator action in response to a turbine trip would be to verify automatic plant responses, such as reactor trip, bus transfer and re-lief valve operation and then proceed with normal plant cocidown using w

th isolation condensers and relief valves if the main condenser is un-available or if the bypass fails. 2.3 Loss of Condenser vacuum ( Te pic xV-3) A loss of condenser vacuum results in a loss of the main heat ' sink for 22" the reactor. A turbine trip occurs at 20" Hg, and a reactor scram at Hg. This event behaves similarly to a turbine trip with bypass f ailure, since the bypass to the condenser is automatically blocked upon re-ceiving the loss of vacuum. Relief valves and the isolation condenser are used to remove decay heat. 2.4 Turbine Pressure Reculator Failure _ ( icfg x v' - 3 ) A steam pressure regulator failure in the direction of decreasing flow is mitigated by the actions of the backup regulator. As pressure in-creases, the backup regulator will take over to control reactor pressure. This event induces a very mild transient on the plant. f

,--M-MN,

_,e uuuseur

~ & c_ _ AMAt yJt3 cl~ -fMC ~11/A6'AIC TAlf KV[A D /$ ,E QUtvALtnsT ~7b C.UKnCd 7* MRC C <t* TEX /A, ~1913 t.s 243ED 04 'TH C ' EC Llo ulld G~ : 6,v,a w x a m v a m s - ~ s.s,s u n o c w i r e m ne r- / Al TIA L CoNotned3 ,4H tsc.c cp TA ot.C (A<.c ur. A ricNAt popCL f3l Ali usco ~ td ~ 7 74 E ANALY t)

TRe~

i4ccgp?MC5 CA't TrictA fon. -rdC G AN 7* i eE -rMcr -SAMd. d 3 f 4 I 4 h 4 -g> em. em a s a s W-. -'6 m,

O 8 Type.VB (8x8) Exxon Nuclear; Fuel i TURBINE TRIP ! FIGURE 3.1.:1 - a OYSTER CREEK ..a... ...... +............ --...s. 175 a.. .. +.. -..... 3 i 'Py4cR LEVEL 4 22. '.-2;T FLOX i i i i i3 ~ 5..,IOC. :Jd, T!0* *c.rLO4

J

~..

s;._ s. e c. E.. p _u s m s. .. c.'.W...:' T.M. . ^W

c..

j ....g.,........... .g....a._ .. g. 1 , $.3-9 .t

j i

..-..-......t... uy.. + . +....... -. -t 5 4 S 5 S -l 5 5 5 3 1 7 3 3, -- - ~. e 2 u _7- ... I. m $/s,.. RQ3,11't ".. e i81f N,. e m c. - e \\. \\ t s i .I 't

M t

g - p.. ......i......... g .... j. J 75.j g .,Nq, t

t

\\ f s. O: e 1 1 l. t

f.

gi .t he = \\/ j. ..... h..... w........ f.. ........4... .. +...... ...}..... l C) \\: 1 .M I i \\ 5 N i N*.9 .1 4 %.=-w % n-o ..3..............+.. 4, 'v ..p. 25. 3...... o a 9

'"*=-ap%%

1 1 i_ 1 1 4 s. ga 4 g o.- .a6.. 3 e. p.. -..r

~ -20 t o g e O ..4..e.... ......6...... ...4 ,,,,g...............p..........................=aq,..~.....ao

4 1

1 g g i. t i g 1 e + . O ? N t . M ....,...............g..................e..l..........~ p....... ........+.............g....

e 9

t. .t l ? ? p

O

..... ~ ....L.............. a j. 1 i. U .+ gg 4 ^ t b en t I H b M Q. O v^ ..d...... .....t.... 'l P")...... W.. ..L.. 4 e4 t

QN 3~-

v C 4 "E.J ? C Utd O La t W X X %8 LAJ O 8.A ft' Lh tal 1 ^ p.ta1 >.. ,I X ' a.7 CQ.... ,....o........ ........ + 6L "I 'I (Q J.J WlaJ CO 448,/4 4 gq q ~~- L e t 's t >y 0 O. 6 O. >= 4 N = u...... -..... -.... M 6 5 L; g b.e Of e ti H M s e. N LAj P M 6 0) CO ec p W e e 3.f....p.. ~ LLJ * ..g........ ...r.. ......-.~ n .. 9 LaJ. gg .t + g 4 i .t L) tt tr2 >= L1 e

o. ~. 7.....

.v ? J3. 4 ? N 3. e ...s..... N p. ..b... ..e... OO ^ g g I LJ t Oc D h O' ti. f4 o to c. Qs 'O o o N4 "4 O C O V O (J V 4 6 I B . = mm = _ ~- - n ~ ~.- -

........-......4...2.1..... ....g......... .4..... o 44 .ai............g....... e 3 4 a ? t.

i..

I O .. -. - + 1 M,4 ..j.......... 2 .... (.. ....j....... .....,......~........e... k. t 'O 41 ...a.....,, I + r* O 3 I L6. 4,4 P W.. ,.4. b.. 3.. V

0 y

H. 6-9 s, "U T a 3 (AJ u.3' X J c cH i t.J >* O v. Cl X .J J ..e.. X e LJ e4 sL).. 7 a.. (c. f

Q UV LJ g

u e4 O

u. I e

v d og O.4 CD L ~ -s C O V-h e I w....., t-s g,,. ~. .h - q~q m \\ p% i b ^ b N, Mw W Ct* %) % - V C' i4 W t-g p. O a ~ I U C 6 g Y,.. ? e M 4 .~ tas Q* e r= g a O 1 w i. ts. k + p 4 o N d G O f Cl) OIIV3 *.%TIJ 154ji 7 *'I !1 00 l..'.64fkl!!i WW

l r FIGURE .2 - a OYSTER CREEK TURBINE TRIP WIT BYPASS Typs VB (8x8) Exx*,n Nuclear Fu21 0 r i.

*-----.-.---4.+-

t -.--e-4- - -- -- ---4.-

l. * ' '

280. --- --4.- .l......-.- 1. 0.:.JER LEV..EL 12, - ei FLur. i . E;IPOA.STION FLO'.4 i ! 0. er.:r.eM..cP'1 FLOW 5 VE".EL n.7 i4 w.

c. l !

e iw s

g ~ ~ a-t * ..g...~t~~~~~~*-**ta-~-aa-*~ g ..m.q.... ...m .9....., g 4 ).,.... f. ? = g ...~....p.........m.. --3........ ...........L..........p..., a 20,3.... t e i + i i 2 i ~~~~~~-~~. ---'~~~a-~~-~~+- \\' ~~ -- -- -- ~ - - ~ -. - - - ~ ~ ~ ~ * - -~ ~~- +-- 1C0 - - - ~ - - ~ ~ ~ ~ ~ 2 . g l N 1 ( ..N ? e g l i u_ l C M l J +O I ..q.. 3 . h, @....j . -) - e3 A4p_ __ .) 4 ( g 0 e 9 3 .f ?..~ -m a ~W a .'/

l

\\". j j \\ U e-N co..L. . "ll r I 6 J v I e . f f

g

? !l l N es t.

t...

..I . 4.... f....... J t I 4Q I g %t - f ; " 's { ) 'g . i s s ? e 1. 's., l 1 I g w.~ . e. 14 ..3 1.s 24 e. e O o

E l l s e O O s. ? a e FigureQ.1.2-b tj 0.YSTER CREEK. TURBIT!E. TRIP WITHu.uT BYPASS Type VB (8x8) Exxon 14cl. ear Fuel 4 ..-......-.-4..-. .. 4 .4.......... VE5 EEL PFES5URE Chit (CE (PSI) i i so.. i 1. j i. i i

2.

VESSE;. '#.TEF. LEVEL (Its.. ) 4 t. t g sca. i j i t i + e 4 ..............j... ......f... --3............ $(@.4 -.. a 4 8 I i -s e .....4,... .. 4 ....-.....-( g..............3.. 3... 12Q.. 4 3

8 4,

} N I ? " bJ 4 6 ? gl i ^ ? .?. _ _ 3

g..

..*.....".t... -t...-. i ...g... $Q q( .\\ i .o s + s. \\ 1 g... -.. . bea 1 = e m. r,- s t-i i 2 . - - w- - - ~ .v -0 a - t i e 1 0 .s. 4 8 In 10 14. 16. 18. 00. 43

24-s' t ...m .at...a m..e. ........a.....-n..........* O e q ..........o..4,......6.......u.u.....a.4....m...a~..a m.ama m. i d I g e + t 6 t i 2 .t t I i. 3 I L 4 i. i g t ? t i I' t 8 I t I C g..4...................t..........,........I... r= i C.J.g p .........t.un.... 3 I I. 6 6 t. t L. r= I, g t. a 8 t. t. t U 3 F. ' . I' .........~.......t.....,.a............ C......................t............ ......... ~....... i l M t x I x b, i f. l g LaJ j. mg t t. 8 i X t 03 s. t

y

......'.........+..-....-..,&.......u.......*..,b......... .4... ..a.~ =. i 80...... O... >4 g i p W . E t t >= 5 t t 3 J 3 D.

W

- 'tL T. 6 e 'h CM l ! u. o p t 6 1

UH t

t i t C i ! JJ t .t

q

.***-*'+....i..* eW p.. I ut. '. ..t...- m W. j m .o t 4 83' xv C I I g W e4 4 e si e CC-aT

t 2 i

f p g g 8 I p. . O.. p... T i +4 N w 1 6 .T. I. .u ,e*. 9' 1% l -(D I C W 4M. i .L....... -.......... >*............ w..c =. .....4.. ...3.... 4.* CD N rf*., .1 M D l t. 6 1 T i N ,f y' I j k,' t + I ...t= .~. -a m. e.. .....,.Q............,...................... ..t........... M i I t. t t.

t..

w ] I. W' ? Cf. t 6 t. 1 W u 3 g l. M g ? i g W H i j m i q >= ? g...f........ 7.. ....y.... j t t 3 I' U j. + 4 N .....4 N [ p......... ,...............h.. .........4..... 6 i ud t M t -/ a t t i r f j Lia. ?. .g. ....t.... ........e........

aa1...u.*-..

9 -.g l (dfL4*tli fM l 6 4 > Ttl 4 1 04 *** a's

1 T. ) t i;is ti t i i

l

'e& -

I - y-y .= g .yy-m -3 r -y 7 ,r,.y, m .-wg- ,7-,,,,.,.,-o- -- g - - = - - - 9 e

e ' l4-Xfl. /5. ', l l k. i

115, 818 0C. Tu%INC 1 RIP WITHOt'T SYPPSS TR3rt 1415 N F'

sr e eTtni 'l t Kat nus 1 1 sccw.tse r s. rt s. g 9, FCitt %fDW9 RSs 5 t. tccurn rim 4 ^- / 150- ~ 121 sl s y 1 s e, - sl t \\ "100 - }/ 5 l s i o { r l 7 L h l M15 HD ~, J l w 'f 4 a e. ~. SU Q \\\\ Q ^ 75 i \\ l Q-a, ~,-;r, . -&.= I i, + 1 .s q ( 0 i 8 12 les 20 2t 25 32 3o 10 D TIME, SCC i r l l r3 L3 tyuj,u 311 O C 1CRb1NC 191P WITHOUT 61F45% FR:!M 14 7T "W .rnt,wmr New 7m t [- 150 vttia ef te veCL 3, b - - - - = i 12s I gg. b l - is Mi ~t==. = LJ. so t .V k a- . a d J ip.j 2s - 'k, \\v/ i \\ f,.. ,V '" " p..-.,%k, ?' ~t 'M f.. qds ,4I lt .g e y .Y JL t

A l.'

2h ~ ~~5' ~ ~ ~ s 12~ 16 iC 21 25 32 34~ 7 0 -- TINC, SCC , -s t w ~ .3 -r -m .y pq ~ - - _m.+wa ++.iw- -y-,.

DRAIT: ' 3/5/85 ~' ~ " ' ' ~ ~ ~ ' " "- o Z.S LOSS OF NOR.'tAL FEEDWATER FLOW ( -ro P i C., XV-S ) A loss of feedwater flow could occur from pump failures, feedwater controller failures or operator error. Loss of feedwater results in a reduction of vessel inventory which causes water level to drop. ~ The icvel drop is terminated by isolation of the steam system. Reactor protection is provided by trips on low and low low water levels. i The Exxon analysis for the loss of feedwater transient was submitted 1 i -rf r :da :b.; t' -':0 P i ; t in ref erence.%. The = e i -* c -a-t j. rr,ne y-t.c c pre": c The cnalysis conservatively assumed full .J power and an initial water level one foot below normal operating l level. T*.e transient is more severe from high power conditions because the rate of reactor vessel decrease is greatest and the amount of stored heat to be dissipated is highest. Having the water level one foot below operating cor.ditions minimizes -he initial system coolant inventory. B d ? ""+!:. C00',......ca 4 i l

4 w 1 4.s e n-m3 -

a m m e g *** 3 *; f,, ;g7- _ 4s * - t 4

1*

e.m a e o *

9 e # '.s _ -,

4 i --w 4 ed M 12 L e,, r,, _ - 3 r., n - n a 4.,-th' _ M, C. P

t t-H - - 4 sc.:,enh All automatic actions are assumed to occur.

The sequence of events is: 3 - Ti!E (Sec) EVENT 0 Loss of Feedwater occurs q 1 3.5 Feedwater flow decreases to =cro ,1 4.5 Reactor Water Level reaches low icvel-setpoint and reactor scram occurs 1 4 i 6 f _ ~_. ._.. -.. _ _ _ ~ - _ mm.- peyi--w g w+ . -p ue wa -w

= * -

e-

i-9

++y .y 9w -w-.c--e--i , - - -,iq-m-y- --p

o. l TIME (SEC) EVENT 15.0 Reactor water icvel reaches low-low level setpoint and the following events occur l . Main. Steam isolation valves (S$IVs) 1. ~ .begin closing (10 second closing b time). 2. Main recirculation pumps trip. 3. Isolation condenser return valves 'J' signaled to open i l-4. Core spray. pumps are signaled to start 35.0 Minimum downcomer water 1cyc1 of 5.36 feet ~ above the top of the active fuel is reached. After MS1V closure, the isolation condenser system initiates system depressurization. The maximum dome pressure during'the transient is 1047 psia, below the setpoint of the relief valves (1065 psia). The ' minimum critical power ratio does not decrease below its initial steady state value. Beyond the first 125 seconds of the transient analyzed above, the sequence is straightforward. The limited assunt of inventory makeup availabic from the control rod drive flow is not expected to raise downcomer icvel at a sufficient rate to cicar the low-low level indi- != cation. Since the various safety systems 4.sve been actuated at this level, no credit is taken for operator intervention. 9 w w u-

e The system will continue to depressurize until core spray flow is introduced to the. vessel at approximately 285 psig. The water level in the core at this point has been calculated to exceed the low-low-Iow level setpoint (4'8" above the active fuel). This level estimation is based upon fully collapsed Icvel of the fluid mass at saturat, ion conditions. Following initiation of core spray the icvel will re-cover and the event terminated. The corrective functions for this event are automatic. The operator performs a monitoring function to verify the automatic actions and attempts to restore feedwater ficv. The operator will manually cycle the isolation condenser operation to maintain a vessel cooldown rate of less than 100 F/ hour when reactor level and pressure are under control. The 1,0F had been previously analyzed without taking credit for the cooling and depressuri:stion ef fects of the isolation condenser. references 8, 9, and 11. In these analyses the relief valves lifted, Calculations show that the core remains well covered after reaching equilibrium at a pressure below the relief valve set point. Hence failure of the isolation condensers, which would be the normal method of cooling, has been analy:cd. Only a complete loss of feedwater incident is, analyzed, since other transients such as a trip and restart of a feedwater pump are less t severe. The consequences of this event do not result in any temper-ature or pressure transient in excess of the criteria for which fuel, pressure or containment arc designed.

on 5/2/79, at.98' percent power, o pressure spike on reactor high pres-sure scram switches caused a reactor scram. A complete loss of feed-water led to decreasing reactor water level. An' operator manually initiated MXIV closure and isolation condenser operation was initiated. However, because _of closed _ recirculation line discharge valves on all five loops, reactor' triple low water level was reached at.about three minutes. ' Reactor level was recovered when a feedwater pump and a re circulation pump were restarted at about-40 minutes. This event is described in reference 27. As a result of this event administrative, controls were established to insure that at least two discharge valves, on recirculation. lines are open to insure. adequate communication between downcocer and core regions. 2.6 Feedwater System Pipe Break (Topic XV-6) These line breaks are considered as a subset of the reciruciation line breaks in section 7.1. e A e e 9 e 4-5W y- ,v-

e r Y 19HAL YJf.3 0 5~ -fM E Lo.SS of PECO4M-TE~A. g vCM ~r /3 . EQUtvA t en-r 'n CU4ttCd 7" d KC_ CKt 7xtlA. pis i.5 24seg OA '77f C. . FoLLo us inr. : - bitttt.sAtt. Wlc2 4917 s!& A S 4~ V N 71 M.5 A 44~ hACC t.Ji^tW / [J PE C 7" ? t rJI T,A L. Co H ot 7; oa f Aa s c.c enrA atc cistc u<.s-riwse p ostcat s<c useo ~ al 774E 14NALY O coi1. 7xc Eve-47' fes vste %;. -707 4CcEDT+MCG CAI Te72ti9 s ~ i D 46 i l 4 6 6 8 9 6 e e ,p. _._M

~

V"= w ^ *, -. . _ ~

a s - ,e e 4 i I O 4 O 1 o Mm o.* Q

VC L.

eo 1. O D M N r" O b CD c,,, ,a g b. b. c i~ Cg y O

t O

Jb -J' a,. O .-4 D _._%..('..,. N u 4 O a. J e e C.*e T* (s d' e 7- .2 s .>> r >- w a o e g C t Q' of.7 J tts L_t <,j,.: I.J D 6 I uf 'cv gI u, g n,J, tt. 1.1. U e ot 3 .e n.w . c u v. w 1 n uwww o s... r o - u bj I., m s o aw _ a, g. u a- - v. g ~ LL. s tt ~ o lo, { y m e t o f .J n 1 l, y-4j w ,e w 3, u. o t fr n -} w ~ %, _- -r o w-1 u 3o J ev / l C' o ~# l/ } w-cu t e t. o e... o o w o o o o o O to -r N O 03188 <!O IN3Td3d r .i.--, -o

0 0

i '

. - i 2 0 t 8 WH 1 3 0 I;' 0 3 5 9 3 1 2 n h 0 r S o s 8 I r e ~ .t l:

i. ;.

[ i 1 f i'. 9 ) 7 4 t- ! r e tM w! W t I f a o. M w 0

.E 6

e i 0 d 3 - V 4 1 e 9 %l 7 1 F 1 i 7 0 3 R 8 f M sE 8 o O ES T s R 'J 0 O s E o F 2 2 S L Ei. iqi iE 1 t P. N sS ~ c E tE C 'r i/ T E u R s _S W 2 s 0 e O . 2 0, r E p 1 E N E M d F n I a T F l O e S / t v 0 e 3'I' i 8 l SO l L e ss K e E ' v E \\. . I 6 0 n i:-..' R i C \\ \\ se E \\ -\\[ . ; ::l

.lj

0 C R gna T \\~ h S Y gl 5sl 4 O 2 1 e rl 0 ru y ,. i. i# j' - 2 q i F \\ t .1 1 ) O 0 6 . N .l i O'I: . 2 C 0 b T 0 0 4 B 1 I 2 2' o ~ s 3

,,.i i'

j,:! . I.' ' 3 . l 2 'i .f- )!:; .je,l,*tI

P!i fif Ib! '

\\

Page 1-1 1 \\ DESCRIPTION OF TRANSIENT-AND 1 SEQUENCE OF EVENTS RELATED TO S0?AM. OF MAY 2,1979, AT OYSTER CREEK NUCLEAR GENE?ATING STAT *0N INITIATING EVENT: On Mayl 27,1979, at 1350~ hcurs ~, an" inadvertent reacter~ high cressure scram - occurred during required surveillance testing en the isciatien cendenser high pressure initiatien switches. Two (2) sensors (RE-03A System I and RE-035 System II) (see Figure 1) of the four reacter high pressure scrs= sensors share a co==cn sensing line with the isoittien condenser high pressure initiaticn switches being tested. The technician performing the test was in the precess of verifying that the sensing line excess ficw check valve V-130-1 was cpen when the scra: cccurred. The scram has been attributed to a accentary simultanecus operation of switches, i RE-03A and RE-033 due to a hydraulic disturbance asscciated with valve manipulaticns The required by procedure to verify the positten of the excess f*.ew check valve. hydraulic disturbance also caused a mcmentary trip of the isolation ccndenser initiaticn switches (REi5A and RE153). These senscrs were net clcsed icng encugh _tc initiate an autematic initiation of the isclation condensers, since a time delay is involved in the initiation logic. However, these sensers also are used in the autcmatic recirculation pump trip logic which did operate in tripping the four cperating recirculating pumps. No autcmatic time delay is involved in this icgic. INITIAL CO.',M TIONS: Plant Para e ers at the Time of the Scram: l Reactor Power 1895 MW: Reacter Water Level 79" Yarway (13'-4" Above the tcp of the active fuel) (See Figure 2 for water level reference taculatien) 5.4' GEMAC ~ \\ l i i l I i ,7

e a - Reacter' Pressure 1020 psig Feedwater Flcw '7.1 x-106 lb:/hr ~ 4 Recirculation Ficw 14.8 x 10 gpm Ecui: ment Out Of Service; Relevent te Even: Secuence: A. One of the two (2) s:artup transfer ers. 53(5ank 6), was cut of service as permitted by Technical Specifica:icns, t perfor an inspection Of its associated 4160 Volt cabling. 551 supplies effsi;e pcwer to one half of the station electrical distribution syste (see Figure 3) when power.is not. available through tne station auxiliary transfer er. The 4160 Volt buses which receive power fr:: 55 are IS and 10. Bus ID su plies pcwer to certain redundant safety systems. Bus 10 is desigr.ed to be ;cwered frc: !2 Diesel Generater in the event pcwer is not available frc: either the auxiliary transfer er or s:ar :: transformer. Bus 13 su:piics a160 Volt pcwer to r.cn-safety related sys:c s and hence, dees n : have a diesel backup power scurce. B. One of the five (5) recirculation loc:s (D) was r.c: in service due to a faulty seal c cler cooling ceil. The pump sucti:n valve was cpen, the discharge vahe was closed, and the discharge valve by: ass valve was cpen. No other systems and/cr cc penents important to the event sequence were out of service. EVENT SECUENCE: (To = 1250) TIME OF EVENT (Sec) EVENT DESC?.*pTION O A reacter scra cc:urred for the reasca previcusly described c upied witn a simultane0us aut0:atic trip i of the fcur Operating Recirculati n ? umps. The Cen:rci I w-p t*--y +er-e++-g-- .w; y ,,y-- 9,,---- m-. 9e, 9g-a-p -ww m p sy,-%w-w7yg-y-yy-,,.y9,y-nyi,w" yw-w -TMv' 'T y

Page 1-3 TLME OF EVE.;T (cent) EVENT DESCRIPTION-(c:::) Roca cperator verified that all centrol reds inserted and preceeded :: drive-in the IRM and SRM Nuclear-Instrumen ::icn. A: this tice, '160 Volt ;cwer was beingt supelied-frca the. auxiliary transf:rcar Auring the coastd:wn of the Turbine Generating System and the' Feedwater System was in cperatien. Recirculation ficw started decreasing due to pure ceas dcun. Steam ficw started decreasing due :: Icss cf heat production (scrad.) but feed ficw remained at the full pcwer flew rate. Reactor vessel pressure decreased tc the pressure regulator se:;cin: as steam flew decreased. Reac::r water level began decreasing due :: steam void collapse I in the cere. 13 The Turbine Genera::r tripped at the no lead trip point which initiates an autecatic transfer of power to the startup transformers. Fewer to Bus 1A ano 1C successfully transferred frem the auxiliary trans-former to the SA (Bank 5) startup transformer. Since SB (Bank 6) was cut of service at this time, power was lost to Buses IS and 10. As designed, Suses 18 and 10 separated through Operatica of breaker 10 and a fast start of Diesel Generator ;o. 2 cccurred Oc ;cwer emergency leads en Bus 10. i' B

Page 1, TIME OF EVENT-(c:nt) EVENT DESCRi?7 ION (c:n:) Less of ;cwer :: Sus 13 resul:ad in less of Feedwater Pumps 5 and C and Condensate Pu::s 3 and C. Al thougn pcwer was available :: the A condensate and feedwa:er , umps, via Bus 1A, the A Feedwater Pump tri;:ed en icw p sucticn pressure. Since water invent:ry was leaving the Reac :r '/essel through the Stea: Sypass Valves :: the Main Condensers and a high capaci:y s urce of high pressure =akeup water was nc available, reac:cr water level and pressure deersased. in additica, the icss of pcwer :: Sus 13 caused the d B Cleanu; System Recirculatien Fump to trip which, in ~ turn, caused an isolatien cf the Cleanup System due to low flew through the cleanu: filter. Further: ore, cne condensate transfer pump ar.d the cperating fuel ; 01 cooling pump tripped. An unsu::essful attempt was made to restart the A feedwater pu=;. (The reasons for th'e restart failure are described later.) (EventRecorder) 13.6 Reactor water level decr. eased :: the Lew level scrat setpoint which is 11'5" above the : p of the active fuel regica. (Event Re: Order) 16.8 The output breaker :n the No. 2 React:r Fre:ection Syste: M.G. Set. tr'; ped due :: less of ;cwer to the drive cter. The cu:pu: vel: age frc :ne M.G. Set had l t i s-

...t. Page 1-5 T TDtE' 0F EVE lT_' (cent)' . EVENT DESC?.:?T:CJi_ (cent) been maintained by flywheel action since the time of the turbine trip. Power to the M.G. Set crive me:cr l's fed indirectly thr: ugh Bus 10 which was deenergi:ed at this' time. 31 The No. 2 Diesel Genera::r Breaker closed and supplied pcwer to the 10 Bus. A seccnd centrol red drive pump -started. ~ 43 Reactor water inven: cry centinued :. decrease due to steam ficw to the cain cendenser. In anticipation of a low Lcw React:r Water Level au:: atic isciatien cf the reactor (which oc:urs at 7'2" above the top of the active fuel regien), a canual reacter isciatien was initiated to censerve invent:ry by cicsing the Main-Steam Isolatien Vaives. This action was taken at an indicated water level of approxitately 30" en the Yarway instrument which corresponds o g'3" abcVe the top of the ac:ive fuel region. It should be noted, hat :ne uccrease in indicated water level and pressure was amplified by the effects of in:r:ducing cold feedwater in:c the vessel during the 13 second peried prior to the Tur::ine Generater Tri;. The cold fecewater reduced the steam voiding.inside the vessel there:y causing a shrink in l t water level. i - a b i -c.-, _.me=- ~ ,-...-a.-. ,nw .,w, -e, a c-v., a., .--w.,

TIME OF E'/ENT (cent) EVENT DESCR!pT!ON (c nt) 49 The Main Steam -Isolati:n Valves fully closed, thus stopping the loss Of water.inven :ry fr:m the vessel thereby causing an increase in reac:Or steam pressure. Indicated reac::r water level started to increase shortly ~ after iscla icn, wnen reac: r cecay hea: ~ re- + established a steam ioid distribution. (EventRecorder) -59.6 The reactor moda swi:ch was transferred frem RUN to REFUEL. 76 (1 min. 16 sec.) To establish a sink for ne removal cf decay hea; frca I the reactor, the 3 iscla:ica c:ndenser.was placed into service. At this time, the Con:r:1 R: m cperater closed the A and E recirculatien 10cp discharge valves (these valves-take appr:xicately two (2) =inutes tc close). It is pcs uiated that at this time, be:h 5 and C loop discharge valves were also closed. The 4 l conclusien tha: the five recircula:icn pump discharge valves were closed is based upon le p tamperature s response later in the event and is further sup;crted 1 by the Lcw Low Lcw level a: 172 seconds. The 0 loop was isolated previcusly. (See the equipment Out of service se::icn). (EventRecorder) 90 (l' min. 30 sec.) The reactor L:w water level alarm cleared due :: the f water added fr:m the iscia icn c ndenser to :ne Primary System. ] i

M~

^ ~~ & *WW _^*

page 1-7' TIME OF E'!ENE (cent) EVENT CESCRI?TIO.l (cent) ~ 96 (1 min. 36 sec.) The 3 isolati0n condenser initia:ica valve fully opened after 20 sec:nds. The temperature of the E - recirculaticn 1000, which serves as the B isclatien condenser water re: urn :ath, decreased due ::~ the e'ffects cf ccid wa:er frca the isciatien c:ndenser. .The 0 recircula ica le:p te :erature did not change appreciably. A, 3, and C r,ecircula:icn lec; temperatures increased slightly. The hea:-up is, attributed to natural circula:i:n thr ugh the partially open discharge valves carrying h: water (536'F) warming the lines previously cooled by the effects cf cold feedwa:ar. The reduced flew area be: ween the icwer dcwnc::er and icwer plenue area, due to the slew closure of the discharge valves, started to cause a shift in water i invent:ry fr: the core area to the upper and icwer down:crer regien. The shif t was due to the isciatien condenser returning c:ndensed steam from the core area to the downe ers. The water inven Ory shift c:ntinued as the discharge valves acved to the full clesed position. (EventRecorder) 172 (2 min. 52 sec.') The reactor Lew Lew Lew water. level instrumen: trip last recorded ::in: en the even: re: Order. point was reached. This was pr:bably caused by the voided mixture in :ne separaters having drained to the upper plenum, causing a redu::icn of static head abcve t the Lew Lew Lew wa:cr level instrument. This does no: 1 l -~. -m-y,

Page I-d ~ ~ '

TIME OF EVE"7-(cent)-

EVENT DESCE!illil (con:)- necessarily indicate an inventcry icss from the core but rather a redistributien cf water and steam voids abcve the care. l 186 sec. (3 min 6 see) .All recirculati:n.1cep dis:narge. valves fully clcsed. At this time, based up n closure initiatica, the cooldewn of the E recirculatica lecp sto;;ed and a heat-up began. The indica ad react:r water level' increased due t: the shift in water inventory. Recir:ulatien 10:ps A, 3, and C centinued : heat u:.. The mechanism cf the heat up_was'due to heat transfer 1 between the hc recirculatica le:p piping and the 4 water in the pip,.ng. Reactor pressure continued to decrease as a result of isolation c:ndenser operati:n. l 250 (4 min 10 sec) S isolatten ccndenser was removed frem service to reduce the rate of cecidewn cf the primary System. . Removal of the c:ndenser caused indicated water level to decrease. The decrease in indicated water level was-due to a return of water to the core regien frem the dcwncemer regten thr:ugn the five (5), two-inch (2") bypass valves areund the recirculatien ic:p discharge valves. During this period, the net water inventory I effect was a st: rage Of water in the recently secured isolatien c:ndenser. The recirculatica 10:p discharge temperatures rea:hed e:uilibrium and folicwed a slew cooldewn trend. I __-s

- rage n-, I j TIME OF EVENT (cent) E' LENT DESCp.!;TICN (cent) L 270 (4 min 20 sec) The reacter pressure increased due :: _the effeces of removing 3 isolatien c:ndenser. The rate of decrease in water level snifted fr:: a ram: cf a: proxima:ely 37 in/ min :: 2 in/=in. The reascn for this change is the isolatica cendenser tu:e assembly was c:eple:ely filled. The fi:w'thr: ugh the five (5) 2" bypass valves centinued, ac::unting for the change in slope. 450 (7 min 30 sec) Both isolation c:ndensers were placed in service. This caused an increase in indicated water level and a decrease in pressure. The A recirculatica loep tempera-ture decreased because cold water frem the A isolatica condenser entered the A recirculatien lecp which is -~ ~ l its return path to the reactor. A portion of ne water passed through the 1000 via its 2" bypass valve, thus causing the cc 1dewn. - 528 (8 min 43 sec) To slow the rate of c:cidewn, the B isolation c:ndenser was removed from service. At this tice, the indicated water level rea:hed a maximum of apprcxima:ely 14.4 L feet above the tcp of the active fuel (SS" en Yarway). This is considered to be abcve ner:al water level fer full power cperation. 'n' hen the 3 isolation condenser was removed frem service, indica:ed water level decreased to 12'8" above the ::p cf the active fuel where it recained until appr:ximately 1212 se :nds wnen A l i N "~ www 7 P' -,--m,y

Page 1-10 ~~ TIME OF EVENT ( nt) EVENT DESCRI?TICN (cen:) .isolatica cencenser was remcved fr:m service. The react:r pressure c:ntinued :: decreass.and all recir-culatien ice; tam:eratures cen:inued to trend downward. Indicated wate-level-was sta:1e at this time because the head of water in the d:wnc:mer regi n was sufficien-to estabitsh ecuilibrium between the water enterin: ~ the core region via the 5 two inch bypass valves and the-cendensed steam returning te the downcemer frcm the isolation c:ndensers. 1 540 (approx) (9 min) The fcur (4) Lew Lew Low water leval indicators were verified lccally :: be belcw their alara setpcint which is 10". The reading appeared :: be at er belcw the instrument's icwer level of detection. 810 (approx) (13 min A recheck of the triale Lew ster level indicators 30sec) showed that the pein ers were active (moving) although 3 they continued to read below their alarm point. The instrument was at er slightly above its icwer level of detecticn. 1212 (20 min I? sec) A isolatien condenser was. remcved fr:= service, thus stepping the removal of invent:ry frem the c re regien. 3 Indicated water level decreased as the water in the downcemer regien f1:wed into the cor>. ~'ien. React:r pressure started :: increase cue to the de: y heat steam producti:n. k ~~~~~7P~~~ ' " ' ~ ..,g-

Page 1-11 l TIME CF E' LENT (cont) E'/EllT CESCRIPTION (cont) 1488 (24 min 23 sec) The isolation condensers were used several more times to control the reactor c cidcwn with cre-dictable increases in indicated wa:er level and reduction in pressure. This moce of operation continued until 1914 seconds. 1914 (31 min 54 sec) In order to more correctly determine the plant cooldown rate C recirculation ' pump was started and the discharge valve was opened. It was noted that.the indicated water level drcpped approximately 3 feet in less than 2 minutes. The C recirculation pump was shutdown and isolated to investigate the reason for the drop in level. In response to the indicated water level drop, additional attempt was made to start the A feedwater pump. The pump failed to start due to a tripped overload on the auxiliary oil pump which is interlocked in the pump starting sequence. The indicated water level started to increase due to the action of the operatiag isolation condenser transferring water to the downcomer region. 'ihen the C recircula-tion loop was started the loo? temperature in-creased frca approximately JCC'F to 470*F. The othe,r recirculation icap temperatures continued to trend dcan. At this time L0w Low Lcw alarm may have cleared. __,n-__ p%wagem o - ~ - * ~.Q ,,g,

r ege i-it . TIME OF EVENT (::nt) EVENT DEICRIFTICN (c:n:) i 22C8 (25 min 48 sec) The A Feedwater pump was successfully started by locally starting the auxiliary oil ;t ; wnich satisfied the required starting interlecks. Indicated water 13'5" -.. level increased to a. level c:rres:ending :: above the t:0 of the a: ive fuel regien. Reali:atica occurred that the indicated water level and ccre water level may not have been the same when it was recognized that the five recir:ulatten 10:p discharge valves were closed. 2340 (39 min 0 sec) The A recirculati:n pt:p was.placed in service at a ficw rate of appr:xt ately 1.9 x 10# spa, thus removing - the dispari y between water level ceasuring systars. The Low Lcw Low water level alares were kncwn to be cleared at this time. Indicated water level dr:; ped approximately three feet to 11'?" above the :Op of the active fuel. The A recirculaticn loop temperatu; rose frem 375*F :: 465'F when it was pl, aced in service. Steps were initiated at this time to bring :he plant to " cold shutdcwn cenditien". 2700 (45 min 0 sec) React:r Protecti:n Systea #2 restored and scram reset. 3600 (1 hr.) The SB transformer was re:urned to service and Buss 13 was energi:ec. b ~

~,.. Page 1-13 REACTCR PARA:'ETERS: Figures 1-4a and 1 ab are a trace of reacter pressure, saturation tenperature, annulus water level, recirculaticn flow, ar.d recirculation loca ter.ceratures frca the tire of the trip to 45. minutes later, when the. tran- ~ sient was over. They are annotated.sith significant events curing the s. period. w e e e N 9 = 8 e 9 9 e =

p i.. 550 1 .s -l.. t t i i - $ 4 0 '.._, Nu --,- ../ t t COMNTER . t.... .coAw .t y0.... [\\q-g. t %j 1 . 1.. ;....6.. t. t....... ,/A g g r

e...%. __ ___

W .l. A, %,t.,. p . ~.. >*== .l. g.. $.=. =, J .g g.* w...=m s ;.:,-- _ f. .7-I.. D g.......!.. 1

f.. 3

.,\\ s .g. e\\ .l. s g .8,.... g 3 \\.,.. I. s ---w---- p.I g ..g.. .. I, $lQ 1 .h ,\\.. '...l........ p qy ..1.

b..

,.3_ ,\\.\\.. t. .1. ... l ..t... t, i . l... .l... .... e,. ..t. 1, .s-g.t t. L,. .5 F 00 L. -l ...1, I.. 4 s ....... l. ...3 I,..l.. s s ..l ..7.. , t. ....a-t, \\ ..t.... 490.. }.. t. . l. e..... .t. i 5 ...L.. .~.

l.....

\\ .t, ................. +.. ....t ......t.... N-A.......... ....l..... 480 .t. i. .t. ....t.. i

i...

N... ...., M,. .t.. 470 1 I,. .. _. i L.. f. t. ,.....i. ...t,......... si. i t. ......t....... ....f. ....t.. l. 4 o.g 4GO .....SIGY..A 3. 1 . t.,.. ..g.... 1.. ..t......... .s... ..........,.......... t C F4 f. t .. =.........................g..... ..t...... r. 9 W et: i.............. n.,0. s.%. N..... p< 450 .\\,, 1, .J,.......... cc .r.. 2 . t... ...I............... . i. w .s.. a n t. p. ............3,. w 44O,. r..

i....

1... ....1.. ..r,................ . i. ..t ........t.. .....t..... t ....3.. 4_0. 1,. .. t,. y ........ =..... ....t.. .t..............,.. ..........i........ ..... _...................J.................. .t. t. .t........t.................... 4,.,.O .......t .i.. ....t........_..................... ............3............ ...t ........[.................t...... .......t.........i...... .t. .t. 410 ..t...... .t.... 4, .......t.t....... .......t.... .....t. t. .....t....... ... t..... ..............................t.........t....... ........._......t... ....... t. 400 . t.... . t................. ..t.......... .t...... ..t....... 39O ., t. .t................. ....t.........t.. .t........... t l.... .............t. ...t.............._.....t.... t............ 380 ..t..,.. ..t......... .t...... s.. e ....t.................. .....t.. t ..t. t ...t t. .1........ t.- ......e......... ,g....._o .8 ..h..e. ......#.......,....8 .g ..9 g, O 1 2 3 4 5 6 7 8 9 10 11 12 13 14 350 - -- s T.---- 1 b kL

e ^ 0 8 , 3 ..e. -f ...g., ..;;.g ,p.. ..i...-.. .g.. .....g............. g ~.. 1.. .....g ..t. ....g...g n.. t.... t..... ..... ;3 .........t.. -.c- ...a .p.

g

$ ~.. . 1........... .t..!. t ~f.... . 7 .. I...........!...,. ..l.. ..l......... ...t.......,.. ..........!.....................1.... s .. ~.. .... ~.............. -.... ............... ~..~.......-................... ,..........1... ..l... .....~...........t....... ............. ~...... ..l.............t.. ...... ~.... + ............v.. ............... ~...... ~. ~..... . ~.......,... ........~.... ~.......... ...~.............................................. ...f................... .......... ~. ...... ~.. .......~............ ...... ~....,.................. -........................... ..l...... ...t.............. .~. .t.... .!.......m.... ...(....1... ...~,.,s,i.. .....l.. .. l...... ....l......... ...1.. l... ....3...... ~ .I.. . - +. - . ~..................... .l.. ...L.,............................... ...l........... 3 ..............,. ~....... -.e.. l,..>... ............+...............ey.... ............................-l.................- t--.... .J. 1 .............. ~.... ............7....... ..~.: .,..,6.................. .t. -....e..... y.. ... 4 .8 . ~. .i... ...b. ....t.. ..o .....n.. ...-............~...I... ... ~..................... -........ ........ ~.. ....-~1. .. +,.......1.. I,.. . ~. ~. ~......... ... ~. . ~...... ..r .......... ~........ ~. ~........ ... ~.. ...t.,.. ...-~..l.-..~....................-* .y-,.... . ~........... ........ ~..... m.. .... - -... ~.... ..1......................... .. ~. -.......................... .~......T.....~..~....~.. ~ .....................o... .............. _..........-.....................l...... .. ~.... ~...... ........S...!.....I............... ........... ~.... ..........,.......!.~...........................~....,...-......... ..~............. ...... ~ --.. ............~.....e.....--...... ...~.... ...............~...-...-...... ..-...... ~ .........~.............~............Z."......~. ........ ~.................. ~ .......X..._... _. - ,....,..........,.3.,. ....................!................ ~.. ~ .................r............-..... -.. ...-.... -.... ~.. ....... -. ~. ~..................... ....... ~.............. .l....1,..... .... ~............. .......................t. ....a........ .. ~.. ~. ~....... .. e ........................ ~..... ........a...................... ..........- Q,.. . ~............ ... ~. ...I...... ...............~.-.... ~........

t..........

.......T,............ ................ ~...... ~.... .........~....... ~....................... .. ~...... ~............ .... ~..... ............ ~..... .......d.......L..,.. ....+'\\... ...f.i........ .................................. ~.......... -.. .~...................... .3.... ..............l.n .......... ~.... ........... ~.. ..... ~....... ............ ~. .,...f. U... a. 3........ ... ~................. yp I r, ,l.. 2*%.... -.. ~. -.. .~.. .....'.%...,..%...b........ - y. _.... ....t....... ... D. 0..i.. dn _. ..... ~....~.. .. ~....... .. ~......

1......

...s........ ... ~.........................,.......... u ............,..........t...............~.... ....... -.....N.... .......,.. ~.... .s........... ~.. ..........._f... .............. ~ ...t... ......... ~............... t......... ...~ L.. ..,... ~ ~.,.... ... ~ -.......... 4. ..w .................. ~.......................... ...-.-.I.., 9...............1....t... o. ........t.......... - .~Q. ..8..... ..t........ . ~..

4.. s...

...'a ~................ .. ~. . 1...... ..... ~...... y ........t...... ,... ~.............. ....V"*..-. . t... ....V... 1 . ~..... ..................l... ............ ~......... ......)%..'%... . w.. .1............. ...........g .. ~........................ -... ........... ~ - -....... ....t.... N.. .g. ...l...... .t.u.... a,f.... l .. v. ..,.....t.e .h..... ,4....... .......... h.. ww.. 2............... ... ~. .~.~..t...... .(.~....... .. ~............ ...l .....l...................... . ~..... i,...... .l. ..%..,,,,a_ ....%.=*%.. m ...~...t. 1 ........... ~.. .~ ......t.... t........

2.......

..t............ -m.... .o.. .t............. 4.......... 4 .........~.!............. .t.. .......r...... .... ~..... .l. .....l...... ......t.......... .....f....~.................... -...... ..~... .-..l................... ....................f..........,. .m. .r.l.~~..-..... .....t... ..m.. J ~....., .. ~.... ..-..)... .l.................... ............ ~. T. y:=:,",.. -... :. :. *.n. -..... .. - ~... -..o .. l..~.. .......-..l.-..~... ...... ~.~. ................. ~. ~..... .~.......r. .,.........g...!,?'a.... ..... ~.

.*.U....

- p.-... .g..a. $ v* + +

1. =

~... . '.. ~.. *. ' '.. -... = .....~.. i .. = - ".. + = ..l......*...'......... * ~.'- ~.. -.. + .....;; N '* <e. ...~o ....... ~. .n + .e. I........... ' + 13 - 14 15 tG 17 18 19 20 21 22 23 24 25 26 27 28 29 30 TIME CGUTES) i. y g-- .-.7 .i--..-+-. y9 4.v g---

O y e es I l e .g ...e -6.0 ] ,.g e,. a.... .,ei. 6 g 4 4-6 9 y o e i 6, 9 .$.a, .j. g i i f t 8S ..g .6.- ..t..6.. p. .g. g.. 9 5 8. .a .....y l. e. {......r..{. . 4 .................t 9 .t .h.-., g 5 ....,.4.... g g t 0..e.. s....u. 6 y .)g .g* 3. . <j .g 6 .g. ..........4.........e.e .. S ..=y 6 . *p..... l. .... 6,.... - =.. *..... we g e..g..... e .a s...-* * * * .e

- - - - 9*

t .. * * ' *.-e 9..,. 6.., .-.*...e p. 4 ...,.... =.* ...a...--

- - .. *. *. =

y_ .6 ee. g g. g*.*

6.... e.. e t

..b ...e.s e ..,5 e .4 ...4 t* ..0 p.. g 4 .s e ... 4 5. g e. g + p 6 9....... ..e4 e 9 t g-.. g .e.. 6 .g .p. e.. .6 e.- g .9 e.e g p 6 p..--,.. ...o..e g. . e... 6 .......g. ..e l... p a ........g 6...... -....e........&e g.. ..8.. e a ..............* * * * * * * * *..............g.......e..= =.. * * * *

g
...p.

te 6 .6

....*....-e..
=. *.

g .I, g o ,.%...e- . e ..4 e -g g-g .B .. yB ...... 4 g............................ 4 ......6........ ,e 4 ..e.f. e... ..,..........e..........,............e. . 6 ....... 4 .e.....e.,...... e.. . ~..... ...*.*9=**..... .e. ..........p .e..... =. ..= ........,......*.=...-* ...***.W** ..b....... ... =. ..e + e.. .wn 6 4.. -t.e .*S .....e. -..e.....a.m.. 4. 9 4 .4...... 9. =..........e. p ..... ~....... .-g. 4.. ..~. .g................. ~-.. -............. . ~........... ....._.....,... ~. ..l,..... .........'E.. . g.... -.. ..../.. .f..... v ........ ~.. ...~.............s--.... - -........ ~. ...../. ....t....-.. ....7. ........... ~................. .w.. .....f....... ....s... ...............o.s.. ....... ~....... . ~.. - -., ..c. ......).. 3 ........~...

u..

........................ ~. -.............. . ~..... -..................... ~... ............. ~. t....................... ~................... ...........~.... ....J................. +. .~. .l..... - -................ .........l... .f............... ...4.%....... p.- ..s......... .C.. ..1....... . ~.... . ~....... ...J.................... .. ~..................... ..~.. ,,r,.-... ...l.................... .._./.... 4.- .................l.. ......1.. ...... ~... . ~... .~. .-3.. ....1.. ..c......... ._..t.... ............1......... _... .....~.... ..w................ .t........ s.;............... ..T.......... ...................... ~............ -. _..... 1 a ~ 4+- w p ../.,,~.. p ....'.-n*****'.s. ..t...........,s.... i ~._ .......,.,............ e .1 g.......~ . --- m ...:~: - -.. .....g. .w .I = N ...1... %.1 ~ i. .i. 1.. . I.. 1, ....N...: J: . i. .....t... ... g ~_. ...t i .,.....l.... - ~, .l. s.. l.... t .,... l

t...

l. .t....... .t ..t. t t

l..

.1 i t .l...... ....}.... .m 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 FIGURE 1-4

1250-16 8

4..

.4. I. r. '-[.. t-n- .. :-BASOLAB s. r .i.. . t..... 2 3 i. ..t. I15 0 I4 7,. CCMPu ER - -- --- 1 i I iDOWN. j' l

l. -

.ii . E' T

t......,..!.. 6. f....B iSOLATICN CCNCENS.ER..........

s,__,_, g ... ()){. - l

s..

..l . :l. .d .:t;. -t. 1050 12 6 :. --...---..-.e.-..'..-----.- [ .l: .8 ISCLATICN

h g[

f ,/ ',< CONDeit_ R. i .i .t. .m ...t.. s .. F.F (.; () ..t:- s,. . _t f

r.,.

t. [

7 .i ..s... e( t, ,s

-~

v .o.. m. - - L - ^.g-. ^ ~ c) t

.. ! ' - - -. - - - l-U 950 - 10

.5' ..l. p u; 2 og - .) a: w

l..

r. au Dm O. u., p

1:

~- m l ou Bc J ,./ 'd a:. c) t. . l.. a w .l...... X d B y. 0 2 6 4 .v. i. a c O O

r..

C) t.4........:... e e ..B C) <c o w

lSOCATION cr; a

w l o e oc. ...C. C.N. O...SC c i. t. ua 750 a: 6 3- .a .i. .1.

t..

t.... ....t.. . r.. .s.

1...........

... =..

t...

. I.. .l. 1 650 4 2 ...............t.. .i 1. .........t.. .l. .s ....... e........... 1 t.. . i. ..t................ _............. i. 550 2 1:.. . I.. -l..- " " " -- --'ir .t.._. ..............l........ i ...t. i.. 2 l ....l.... ,. 7) O I 2 3 4 5 6 7 8 9 10

. y ..a,.... ..l........ 1. ..t,. . 6.- ...t,.... wB. lS L, ATIO?LCOEc.NSEF r-t.. t.. t...

1. l..

. l, 4. eE. ...w ...-.. e .......i.........:.... . a. .....t..............._..... .t 1 y .1 ......t.............. -.~ ...t...................t.... 3.. ...._..t............ ............._t................... ....i m._. .L............... ....._ -............r...... _........_....... _...... .t.

i..

..-S.,C....3 ..t.......t.._. .t.. ... _t.. .w......__. i. _ _... t. .......i t.. .....__D.O.!.'. M.._. 7..... _......_........ ..t.... -~ .........i t... _............................. .B TH..f .__...t............._....__.._.._. t.... .t. 50L m e. ..t.........,...._....__.._._.._.t._.._............._........._..._....._. .,........t.....t..._......_.....__. t. .._. _.......t. ._._..t ..C. Ou.D.e.meR.. ...._....._....... _.... _ __-..........1.._....._.._...t._ l...... ........t. .....__t...... .s..__......t...._....... _ _. ..._......t......_........ ..........~ __._._...n,........ _. _... _ _ .........t........._.+.._....t.. ........t......................t.. ...................................t............. ................r..._..._.... _ _. _.._._. _ ..t, ......_. ~... _ ,.......,..._..._.t...... ..t g i ..... 7.. ........t....................... .._..__.t... ..m ........t.............._,.......... ...........__t..._..-.........._..._t._........ ...........t. ....t.. t. _..... _....._......t............... _............ 4.. a. ........t,......_...... .........t ...................t.. .t._._... ._...t. _.....t.. .t... ..t....._....... _......._....... _.......... .t .t................................... .........t....... ...t.. t ..t.................... _.. ... ~......... = .t. .t. .t .t.... ..t............_.._............_.............. ..__............t........ t...... ...t... _........... t.. .....t,... 7 8 9 to il r2 13 14 15 16 17 18 19 2C l6 s. -l i I i i i I

,w<3-y---

_w ,y p e - - - ~

,e-e ,e* . se ............'1.7'..C...... L 4 (.......... + i .4.... .............+.. .l. .......,.........>........., ~.... ...,.,.,., a... + .......................g.........,............................... 6.......................,................e................... .......I... ...I... . _............. - ~..... .I.. ...t. i .....).* ............i...... .I.... .l.... ..~..... . ~,....... .......y... ...............a.. ......\\..,... - .......}.. s........ s. .L..... e.. .. i .......s. ..........i .1. o..... -. n-e..,..s. CGq.. r..h-e. H. 6 ct G,e....uT iG e.1, ..I.e........ e*

l....

C.* ay 7}... G.-{tG.du,., L .G.r.. s.. + .v as.- n . A c... ......l... g ..:.4.. vT6 l... ....t. 4.. ......s.6. ....I...... p ......,... _....... ~.....p......... .....I...... ....,.{, g. 4 ,.. I..,.... .. ~ .6.. ~... ....s ........,.......................u.. s. 1,.... ..~........~.... . ~ ~.. .1... .....I. ..-6...--. ...{. o......... . v....... 6.. . ~...... 6......................... .... w... ...+........ .l. ~. ..-r... .6. ...............,l.. ....+..

6.,..... -

-= ....I... ................f..............~....... .....s..-.... 1... ~............... ...........7..._... ................,.......l.........l.. _..... ....... ~........... .g.. 4......- ...,.........-........ -... ~. -..,.........,...... ... ~... .w......................... ~..... v. ............~.,. ... -...............!.......,6...... .l...... }... ...\\ .~..+.... ...........L..... ........ ~.. ~ .~. .......... ~. .-..4. .6 a .. ~.... .... ~.. ... ~.............. ...... ~....._............... .~. ~... ..... L......._-.............. ,.. -... ~. ~..... ...c. 'b.........................,.... ~........... ..M ................... + .L.M L../'n..^ r. v... ... b.... {..... w .. ~,............ . ~................. ..,..~............. Q,.. .u... _.......... .....~....... ... ~.......,._ . t V....... ~.. ~........... -................... 4 ...r ........... -.~..... C.- ..~.................. -............ w 4.,.. w ............................_....^^s212'.'*..pc..,.......l....,.. .6._.. .....~. ..T........,...................--. ...M.................. .. M...,; - 4.. ........... ~.. ...... ~................,.................. ..... ~..... ... ~...... .!,.........l. ......... ~................................. . Q,f........ .. \\ i ............ - >....,.. ~....... .~. _ -.. ..,m_ ......... ~ _. w -.,...... 1... ................................................+............. ..... d. ..~....... ~........... ._.5 .......... r...... m. -. -... -.. l..........[....~... .t_..... ..~. =. -........ 6....... ~... ....a.......,...... __-................. ..[. .......,.. _....[. .w....._............l................. ........ o............ .f.. . ~.. .4..... ...............i-....A. .....t. 6. ..................l ,_....f........................- .4... .f..s n.__ .,.......I..... .............t I... ... _.....t... ....... L..... 6.. g.. (... ........s. ~........ s..........s. ......... _..I... . ~.. = 5 i ..]. .............h, p.... ....l................... ..t.......l............... t ....p......... ............k.... t... 1 ..............................u... ~_ ~.. - ...... -.......... ~. ......l................................_....p.....,.......~. . ~.

n-~.

............,.........................r..,.... .............,.....$.........,....e.. .......-.. ~ ~..I.. .....~.......... .-................ ~. ......l. .- ~............ ... ~.....!.......... 2.a....... 9...I -+-- ..l.,........... ~............. .6...... .. ~.. ~ ~.~ ~.....~ ~.. .... ~...............,..... ~. ~ ..*6.~.-......-~~~.. o..... -. + " ~... ~ ~.. . ~.. ...l i..o... 6.... ...,.................................T......................,. .... ~ ..n.................... ..*.***... ~ ....- -...... ~...... ~... ~ ~....... ............ ~.... -. m l.................................. ,.+ .s... . ~. ~ . - ~........,.. .........s.*-.... f.'........-......... .l.. . 6.~. [- - ~.... ..............++ ~... ~-.........*.... ..{f....... +...... ..l ..................e.. ..s.... .....~..I,.

i. u.....

.6 .. ~.. - -. ~ ~p."'".6 ...._t. .-...+"a...n".......~..t...

""t-*

9....... .~..g e_......... 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 TIME (MINUTES) i 1 i _g.

a. L.. '"A RECRCU ATION - ........... ~. / t...... ~....., 4 . ' :.. :... e . :1,:... CTECiRCUCAT10NiPUW F,. 7 1.. t.l....... ....t...... ~ 1, ,t....... . if. .......t........... 4... ...................... _.......,......_. 7.... .........~.a _.......t.............. _..................... .... t. . m .. _.._. 2...... .......t. t.. ................ t... _.......... . _... t. ......_.._..s.......t....... . _........._....t.........,.....t.......,...._._.....r........._...._........_.............. t.. ...t._............ 1 .... ~.. _ _...._......_..y..._..

t.........

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

1....._.......

...t..... .__..t._........ _... .,a.......... .....................t................. _.. ..~_........ -.. _....t _........................ A.... _ ........_.............._=..... ........w..._.._. _......_......_.._............................ ....t._..._.._................ ...t......

t. ~........

.........t _........ _....tn.. _..........._. ._.._ _.. _......_..t.. .......t..._.t..._.-..._.. _..._.....t.. _..... _....... ~. ...........u. ..t. = ..._.._..t......... _. .1........_..4...._.. ._.t.............. .t ....t... .....4.. r.... ~.... ~.....t....... .t.. -. ~... ....t.... t.:.... ... ~........... _....... ....t_. ...................t......................... t... 4 ...................t.. .........t.. t.. .r.... ......... ~..... ..t. ....t. ............t... .t... .....t... _.....t.. .. _............t. ......t........ _..... .t.......,..

r...

._...........t........... ......s . e. 9 t... o i. . t. .e... 32 33 34 35 36 37 38 39 40 41 42 43 44 45 l. FIGURE 1-4B

(..-- z-. 3.0 CROUP III EVENTS' Group III events are' infrequent or limiting events with a low pro-pability of occurrence. These events are due to steam system piping failures. 3.1. Steam Line Break Inside the Drvwell These line. breaks are' considered as a subset of the recirculation. line. breaks in section 7.1. 3.2 1 Steam Line Break Outside the Drywell - - - - s- ~ A steam line b'reak results in a sharp. increase in steam flow Jand system depressurization. The flow limiters in the main steam line i limit the blowdown rate to 2007. of rated steam flow. The turbine admission valves are closed by the controller in response to the decreasing steam pressure. The main steam isolation valves begin to close within 0.5 see in response to high steam flow or high' ~ pipe tunnel temperature. The maxi. mum 10 second valve closure time is considered. i : A reactor scram occurs upon 107. closure of the MSIV's, with low water level trip as a backup means of protection. j. i Closure of the MSIV's halts the blowdown. The analysis shows that the ) core remains covered, even assuming no makeup flow from feedwater or control rod drive systems. Further, the amount of fuel damage prior to isolation is small so that the' radiological consequences are minimal. 1 i l f t y .w _._m_. y-- _.ma_+ m -y- -m m --y_. w

- j;. ,,/ This' event is analyzed in ths FSAR and the radio og cal effects.vare ~ l i later scaled up to 1 30 Mwt..Since the plant response to a break J- .outside the drywell is not sensitive reload fuct characteristics-the-analysis is considered acceptable for reloads. 3.3^ - Radiological Consequences-of a Steam Line: Break Outside Containment' L The radiological consequences for this event were calculated using'the~ ~ r . criteria.of. regulatory guide 1.5.in reference 29. The calculation used the limit for primary coolant activity concentrations and the maximun closing time for main steam isolation valves as specified in'the technical spceifications. The results of that calculation are listed' l below: Site Boundary (414m) _ 2 hours) ( l-Whole Body 0.17 Rem. 13.6 Rem Thyroid Low Population Distance (3,218m) i (30 days) +~ Whole Body 0.027 Rcm 2.18 Rem l Thyroid Although these results are higher than those given in. reference 5 the consequences are still limited to a small fraction.(less than or equal to 107.) of the 10CFR part 100 exposure guidelines. The analysis of the main steam line break accident depends on the operating thermal-hydraulic parameters of the overall reactor such as the pressure,'and the overall f actors af fecting the consequences, such .as primary coolant' activity. 'The primary coolant activity is assumed r i ~ --O c --,c_ . + 4,-- m

( to be at t.se limiting value stated in the Techneial Specifications. Insertion of 8X8 reload fuci vill not change any of these parameters so the results of the analysis discussed above vill not change. e 9 O e G G e 4 e 1 ' = - r & --

. m.o v. v. o-o.. 4.0 CROUP IV EVENTS Croup IV events involve a loss of se power. Loss of power to auxiliaries occurs with moderate frequency; a complete loss of ac power occurs infre-quently. 4.1 LOSS OF AUXILIARY Pot?ER [TcPIC gV-4) A loss of auxiliary power could occur due to electrical power distribution malfunctions. A reactor trip would occur unnn loss of AC power to the reactor protection system. Loss of auxiliary power causes loss of condenser cooling water, trip of feedwater pumps and trip of the reciruelation pueps. Turbine trip and reactor trip ensue. The bypass is assumed to be availabic for 1.5 seconds, reducing the power / pressure transient so that the transient is less severe than the turbine trip with bypass failure. The bypass valves trip shut when the . main condenser vacuum reaches 10 inches Hg. Reactor operating experience has shown that vacuum does not drop below the 10 inch ser '-t until after 1.5 seconds. The relief valves and isolation condenser would be availabic for decay heat removal. The diesel generators would be availabic to supply c=crgency power with a loss of offsite power. The diesels automatically start upon opening of the breakers. A control rod drive hydraulic pump, powered by the diesel, can supply 110 gpm makeup flow to the reactor. However, ( i l l 9N--

th9 analysis.shows that even without this makeup flew the core remains weil covered. This event was analyzed in, reference 5. As discussed in reference 13, this event causes a less severe isolation than a turbine trip without bypass, since the bypass is assumed to function-immediately after the trip. 4.2 LOSS OF ALL AC POWER (STATION BLACKOUT) (TOPIC XV-24) This event is being considered as a generic item. No licensing position has been established; therefore this topic is not being addressed in the SEP. m 6 h 6 e e e e m _n

a . e. 5.0-CROUP V EVENTS These events involve a decrease in reactor coolant system flow rate. I A reactor coolant pump rotor seizure occurs infrequently, and a loss of flow because of a loss of power occurs with moderate frequency. [ t 5.1-PUMP TRIP _ (Topic XV-7) 4 A loss of reactor coolant flow can result from loss of power to the. pump, failure of drive motor connections, M-C set breakers or pump failure. The decreasing core flow causes a core heatup due to the I flow-power mismatch. 9 The increased void formatica inserts negative reactivity to drop power back to a level compatible with the lower flow. No reactor trips occur due to the decreased flow. If the loss of flow was due to a loss of power, a scram may occur due to loss of power to the reactor pro-j tection system. The MCPR decreases, but does not reach the limit. 1 l A loss of auxiliary power can cause all five M-G set drive motors to stop, leading to a five pump flow coastdown. One, two or three recirculation punps could be los't by a f ailure of drive motor connections of the M-C sets to the buses. The coastdown from these events is less severe than for the 5 pump trip. A AC N r-art. 6 5 Pun d M as p i s o,o g (ToAc XV 'l) 5.2 PUMP STALL The seizure of one recirculation pump is also connidered. The flow 9 V l .~,w- ~ -- - ,,a r, .s w-- ,---e si. --,wr ,z sv. ,,y.,e..we==- .,er-wg a w - gm* --.-eog iy ww-gg-,g-

~ sere. k reduction results in increased core enthalpy, and a reduction in Mttith". Reverse flow begins through the pump with the shaft seizure. The slow-power mismatch is slightly more severe than that due to a five pump 7%PR. trip, but the results are still acceptable since the NetM does not ( s drop below the limits. No reactor trip occurs. Reactor power decreases in. response to the reduced reciruciation flow. s The loss of flow events have been analyzed by Exxon'in references 9, 11 and 14. The results-show that these events leave more margin to thermal limits than other transients. Oyster Crcck f.s permitted by technical specifications to operate at full power with four recirculation pumps. The effect of a one pump stall from 4 pump operation was evaluated in reference 28. While the. _. one-of-four pump stall is slightly more severc than the one-of-five pump stall, there is still considerable margin to thermal limits. 2 e 9 1 l i i 4 l r

6 t ,e* e ee 5.5 P. v m isf7 a ta X-c A p.v m.fy..iwl on G(m }e ys=1 A., p 6,~t c au ca. A na.p g, ,w,, syv x.x + y . ~.. ~9

p.c c f.t A c L, ~!,cdys v -f u

.hm a nG. you ,o-v 4u, > .e l' l.:K4

u. <w.=-

4+ .m-w< po~p A-i<~. a-.-sp < w. s a,4xuyk--1/i ,c,, ym -i - rk-ed >&, .i$s h y pe 7h<., cc.: t-J t. ,o ld. f.c ~ d> 4,c. J c4 ,.t & & -x.cf c M N // O ef !

),

GN --f'WMk ,/ UC 'C' n p 'bn'c4~da c.et) ,,. W! QW

A:d

,cccme i ./ u n :. s i: ~ ,q) a q -f.-

b h)

N$h s D'. i ){.Q~s id /,, x $<,7{c O,> U 1c..LT' ',*wju.,cmi- / b. ,m./, v / f&~P (L, JX < _N_ - @e

.o. e.. Wd p re t c h a f d M A '

A 4:oL M E vCA *T3 pqMAL YJr.$

of. -rHE i AM CuencuT* M K c._ CKtTEttA. -7.lis s.s /34 Jeu j+ E Q v w A<.en T 04- -TM C Fo LLo ul t*J Cr- : - ' bittaL.AA. WXR fA '72 /s. i4 S CuN Tsca.$ fli A.40C s.JITH ACJ#EC 7~ / *JI TIA L.- C.e N o t T36 4 3 kH

% C.C CPTA 6LE 09LCUL A'7/0HAL

,/4COCL Dl All' uss"Q k til -TH E ANALy 0 ~1EE" 4CCED?MCG CKt1*Cscit9 fo/2. -rHC GVe~^J 7* f CE -r'rt er SA M d. f se 'I, i i i 4 m 1 1 t 9 1 ,,n ....-n r -,.. ~ - -., - -,

p Typ2 Va (8x8) Exxon. Nuclear Fuel 1: d . FI GU ru. 3. 2 a OYSTERCREEK$- TRIP 0F FIVE REC 1 .LATION PUMPS 175.. . i - .......i...............i..... ......i. .......i. 4 i 1.

i

+ C 0'.'EC: LEVEL i 2. W:MT FLt.c' i 3. EE.'!ECU;ATION FLOW ~ i d

/20iEL STE6t: FLCtl e

i4 8 i5. fE L:*'?E? FLOW + ~ e.,.) f...... 8 y + -.~.. ~..~ ~ ~ ~.4 125 ....-......~... - --.+ ~. .. ~. ~ ... ~... ~ _ ~.. ~....... * ~......... l 1 3 3 1 .....,',.....,,,,.,,.1,...,............,.......,..j. ? .. g. 103 ; 2j

4 4

g ~ 4 w N ,e 1 O v.N,, 1. g w N. C \\ r ? La. t. 4 + \\ g ......g.. o 7.... 3.... .4 ...........g... ...v. 7C,.. V p 4 $~J \\ 4 N _5 s u g 5 g s N. 4 - g q 1 I '4 ,g .. 4.. .. 4. g. n.n.r,:.- 4 .r $Q 4 _o4 1 l -aa. C q s S 2$. ? s t 6 i ? .~ 3 ? O. 2 4 6 5 10 12 14 16. 18 20-Q 3r.. e. g,- t

.-m. i I-1 I i I TRIP OF.FIVE RECIRC.ULAT10ii PUMPS Type VB (8x8) Exxqn Nuclcar Fue [ FIGURE 3.2.1 - b OYSTER CREEK. -.~~.9....--.-.-~t .t. ...:...~.. f. .i... ....t .p -. ! 1. VE5SEL PRES 5URE CHANC-E (P5D i 75 l2

vE55EL WMER LEVEL (Itt. )

~. f ~ ~ .y.... $Q..,... g. ..g. t . 2 2

2 2

o p n g p = +. ' - ~ - :" - i-j- 4-25 ' i. ~-*~m~ -~~-I --;-~- ~ ". - ~ ~ g ~- -0 ' w, i 1 w i. e 1 .a + + - - - - - r e- - - - -- - - + - 1

r. -

-25. t I i a + i . g .4 i : ............7 -.7... ..~....;o l I . i .i_ t ...-..f.. .e,Q.. a 4 ( 6

4. -

I

a---

e 4 ~~ - ~ - - 6 2 4 v, G 10 12. 14 H 28. N -100 - 4

l l l l l 3 i e e '^ 3.. 'l FIGURE.2!1 - c OYSTER CREEK! TRIP OF:!FIVE RECIRCULATION PUMPS Type VBi.(8x8) Exxoni Nuckar Fuil --~.u. .- - ~ - i -~ ~ ~ ~ ~ ~~~~ t : a ~ ! 1. FFTP0x. H-L COM ELATIdM ~ 7. s.. j-l

2.

>.It-1 COEF.EL6T10:t 4 5 h,...... ~N.3 i l ...I..

7...

..-~ y .....g............e.... 4 6 I .N i r,o rI. ~ i i# .... r. r.. ~ i 6 e ru- ..~.. - -. i 5 e ,a

f.. y 4

,4 et .o i i fyr 6 .2 4 4 ..=~~.~.4..~**~.g..**~. 1- ......4 F.. 4 e u r J u u I .4 p i

4 a

--~e-- - ~ - ~ - - -+--~~~*-~~s.'- - ~ - ' -~~~-t.- u r-3 ~s q-o 4 g 4 v v.4 ~ e 4 --;~- P 2 1 s 4 .........4 ....}.. 6... 1 ? ...s 9 10 12. 14

16. -

19, 20. 3 6., tit..:, eEe. O. 2. 4.

8 i i O. Type VB (858) Exx:n Nuh1 car Fuel ! l.

FIGU,. 3.2.i2-a OiSTERCREEK ~

TRIP 0FOilERECIRCIATIONPWP i - -- - t. ~ ~ ~~ ~ ~ ~ &- 175. s... . i ..s... . ~.. 52. WEAT FLUX [., .i1 _ POX R LEVEL I 3. 52:ISCJLATIC f FLOW i i4 VIii.E.L 5TEMI FLG1

5.

c c. :r...:.y c e p. c.a I 4

.e~

?. " -* - ~* * " !. " -- - ...i -. - i-c-- 150 -.- t. 3 4 6 .....+.. ++*a- ~~~a - * ~ *

125..j..

...~..a i i + .i 6 4 103 -~1,2,, a c - {" ~ ~ ~ ~ ~ ~~ ' !" - ~ - ' '- ~-- ! "' ~ ;~ *"" '-~ ~ 1 " - -m ~4 L i- 'I ~~~ -' ' ~ * * - - ~ ~- ~- ._,~~ - * %. 4 I m h O -gg 4 d NM 4 i s 5 6 7_ t.s s W - 3,, h-c. ? 2 o 1 e e- = m.. 75..;... ~% p-- -e J 2 w c l i- - + 50 + - l s i l 1 i 1 -- ~ ' a ~ a - - ~ ~ - ~ l-25 - - + r.,- -- ~-+- (~,.n '..

i..

i i. d 4

.}

? Q 0 2. 4 6. 3 10. 12. 14 16. 19. 20. I ?..,.. :. c. er.-

n O 17 'f o.2.8-b OYSTER CREEK ! TRIP OF ONE RECIRCtmJION PUMP !TypeVB(Ox'8-)ExxonNuclearFuel 7s IGUR .i1. VEMEL 'FPE550PE CHA!CE (PSI) i i i. 2 W.35EL &#$TER LEVEL (Ifi. A i2. j ...s e,a s i ~ 2 2 2 ? 7 ? 2 2 2 t .......r.. ..... -.... =... - I 2f, .q... e i l g e '. -~ "~ ~ - ~~ .~- - s -* ~O =g- ~ ~ -~6-- ~ - a i ~ u a i e 1 I O 1 t .1 n i. t.. ~ ...g... ..j. 25... 5 a r a a i .r,a... ,i I 1 i. {.. .3 i .;PS.. 1 [ v. a <r i,. 1 L s - ~ . icy.. C. 2. 4, 6 6 10. 12. 14 16, 18 20. y..vc.. ce.

{ ,,,.4"..9......-. i..................-... =... .........k... 1. o e t ', e ? te ee 1 1 il 3 i i O,. 9..... . N ? = v ....q................... ...... 9.... ...g.- 9............... .t d 3 4 (' -F & f, l g u L&. 4 L: e D H$ j W. e 80 ."0...., w a ( 3* Z. ? .i C ^. O L x4 x i - us N e. ........y.... ..................t...... CO..... g. K H Q. DI e .J CO W a: f it C 3 CH U: E Pl-N CL

h:

J '1* 3 ? ..*P. ' l g .H.....:p.+ 1 u '. - tw g teO ).* 8.) 0". O.4 Ae* 0. i Q. (s. N il 6 N O ,,.g .2... O +-* : -1 3. ),!, I i g D.* U' W LaJ cc N p. g

a

?. .....b.. Z.'. .. k., 4.. O e e O e O., M ' ct: H t s 44 ..g............ l -hf LJ ~ LaJ Cr. LJ e \\ 1 N. I 6 LaJ. H a e ...p..... . 4.. N.... .. 4.. i O ) U l ( N. N N' a M. 4 - o O. 6 g e .....9... ...g... g (2!.ACW) OIWd XTnd 1U3H 7UJI11:13 uf:UIHI!! . s e-.

v .Ii O. O. r. i i, i i OYSTER CREEK - RECIRCULATION PUSP STALL Type' VB (8x8) Exxon Nuc1cariFuel 1 ....... -... ~.~ 4...--'.- ~.-.%.- { l FIGU E 3.2.3 - a - -t-PCC.E.O LE\\. 'EL 175-.

,1.

FcVA. a et-i i 3. EE F;i.'LAT!ON FLOLt !4 LEifEL STEici.F.L0t!

t.'.

C :

r. %.1 L* G C. (.

t ra... .u w a.~. 3 u..... ~t... ~...... .....q.. o4-u g .... g.. 1$.). 2.*-* i 3 .g 8 q j 1 3 12$.. 4 j I ,\\ ....s... 4............ ..........................4..........-.. l g 4 4 t 106).. r e i + u s.> ^ \\ 2 5 ~ os o g q i S g- \\s ? A C f 4 ? _. 9, tpA A _ i ?.s 34- 't. 3. b \\ 75 e- _ _t /

i s=-~--

m c 1- \\ N. w e-- ~ W y aJ i. .7 50.i".. 1 3 g-y ... ~.. t .I l .....i... ..3.. 25 ; .. 1......... 4 l .s .t ? ..... ~........ D 0 2 4, 6 Jo. 12 14 16. 19. 2

c - :O 1

0...3.2:3.3 oys1tacatt<= acciacutationgusto.s..ttn... ryge:va c8x83 exxco socicoriguei j

i-FIG i 1. VESSEL PRE 55tFE CattGE (PSI) ) 7., j VE35EL 6m E& LEVEL (IH.) i 2. = e f --.-.r..--.-.-.-----.s.--

t.. -

n.- ?. s- ? -e-c,0 g 4 I . 2 2 8 7 2 2 o. 2 n 4 e m,5 .?..... .. ~. I e ') t t i i ...g....... 1 ...s... .Q t 4 sa u,) 4 l r. .-p-- - - f. ....+e ............t..

t
t

.i 23., e d.. ..y.. 1 . 1 4 1 -t-.. r.- 1-e -50 f. e> e .....,w...... ~... ...s.... ......s. a.. j 75.. - l + 1, b e + . - ~ ' - .--..~2 ., _. ;.. -----~..j. L 0. 2. 4 6. 5 10 12. 14. 16. 18. N 1 10Ci. T."..E. .4 F '". I

e y.4...............I...........-.....;...~...........p.... -,......-4... g e - e j j i ? s S. s. ,e l 1 I g g I .A O..*. w 1 6 m,4 p ....s............... .....g................., i. t

t N.

F: W d th 3: 4.... .......... - 4 t L... e? W. e e .?

- = 1
l ei y.

3 1 2: [ 1 cv

+

c; ...4....... ...s.... ..............?.- . y.i... g.. O r x: >i 6 >= W ** .C g 4.. ^i W CO : 0 2,* x. ~... CO ! C'>e l 4.. %.# : O>* C .N

e4 CO : gg

}... s., N,4 ,y .ke... 1 w.. ..i.- i a 68 : -Q C:. : Xv 1' h; O p* u; e4

g, h.. e.

m. 4X J:.. t .,n J3 e4 N .. e

c.*

M e E ^^^. t. W V8 1,,1 I a ?. i be .N Z r O. .I b.4... e 0:. ..g.... .s Os 4 (,,,J 6 n* w g g i L) .a w ib ? M e ...g.. ...............+p e- ..p... I M s' g ww: y t,,) 4 CI* ? LLJ ...1.. P=

?

(/). C3

== 1 1 g 0 .......j... ...s.. 4 N .M..n.. r (Q. .4... -g ? 4 t i e 1 m i m c). 6 i LL. 4 g.. ..q.. O ....g.. h. to in ,t c) N v4 G G 3 CtD OIiUd M h 193H W I11 M Wil!MI14 9. M _ w em

m

-a-

l FI

3. 2.!4-a OYSTER CREEK :

FLOW C0t{ TROLLER LFuriCTIO?i (Zedo Flow Demaisd) Type Vti (8x8) Exx~:r uclear .3 ... pg g j- + 175 9.. 9..... i 1. FC:ER LEVEL i 2. 6 TAT FLU): L ? 3. ATC!?CUtATION FLOW !4 ViiCEL STEAM FLOM ~ 5 Ft 0:i4Tio FLO.4 I .l 150.. ~..;- .....~. ; ~ ~ ;..... . 4.. 1 . s. l i l 2 .ll i' +.. - - 12s +. 3 ? i 1, i-2,4e 103 - 4..... 4 4 f' p %.. w 4 g - i U., 4 U ,~ ~ ~- _ 3g 9 3, 3, n G 's i - A .c __ n m _-2 e _ _m-~ 9 u s u s l 7 t I. go .4.. 4 1 \\.. i.- 6 4 25 \\ 4 i 4 O 0 2.' 4 6. C. 10. 12. '14. 16. 18. 2\\ TIME. SCC

...4...............'.4................,......... N 9.f ri.......... N: 3 . se 3: W t 1 4 b - g. i e= i* N ?

v4 g

..............4 4

1 M:

J s ?. e 4 m N j M,

v. I

...4... "O g3 :, ....................L.. H u... D3 ' I g ' ?> I s G, c2. : b': ^ M p m N g gr e 4 .y v A. ..s.... ..s... .... +......- g m.o... y g..

es C". : ** V 7

! Y ? .- xa

u p.

O. W LJ l n'.J X D eO' 4.W L h ,.e g eg .p.. t d >=.. m . g er. ...c... ..g.. et s .O L ! .J J %d- , 3,9 c: tmd N *. V *.A -e l l aj . t.f y e Oi N M: ! e4 N O-=..l. w4 R: LA,,, ! %d 4 J Us q:

2..-*

.i t.J - r. . uJ : N M: w H Ji g . O

* ' * * * ~

J* 0) 1 CE I t.: e. CD.: U:

t.

4 Ji N 4 LL. * + ... g... %e. !. t. uJ : us. i CC : O! 4 M.: N 4 k.J 9 e p j . e.. i C) : 3 .O : It N + ...t.*... g g Ni i M: O*. ' tJ :: t i W: i D. l { oe :: N p g.l %A.. ..h. o,* ..g.. ='- tn o m O M O La O O O b. "8 is a N e 3 4 4 [ e. _.*W p " 6.P*'D'hH 84P'* W-* W 6 - W-g-. I &*M.*.'*9w

W M. N F %u sv*'

I I + 3' 8 ....$......e..........g....................................-......... 9..i...............p........................... 1 F. W.

u..,

i. -43 g. 8 3 t e L

1 g

? i .h-O w 4-e.... ..,..................e...............e..... ..e. I c: O. 3 >C I 3 M' 4 W: - I i .n s ej 4 1 03. 'O x cQ..:......... ....4............. %d i a g. i I 4 W. CL. l 1

9*

ry H 4 ir. ? ...O.. .....4,.....................4... --...k.....,.. . g H e: H g - >= 6 'I g e2 9 w a h j LL I.* u' Q OH a* O

- g. U.>I NY o. fw o.

J CL

1

..........g.. ?. .............q. ....y. 7, g. 8

": u L

O o .~ { r,.s D:A N.

\\ d.

I v ; it 's k ( E ,s. o

e.

N ........* st em...... y WC g. W p H W ,,,J .I to es et 6 uJ g.... t g C fE 6 ? pt 'OO

1 h

6: i J ...{.. ...,.............. 9 ......i.... 4 Q...;.. b I W. W C2: j O 6 ( b: ...7.. ..l, 4.,,,. ,t p... V3 - l' O* t. U '. h1 h c, ....~........a... j. c 4 l i gg 8 O..M. 1 W l M. n. S O e. LL. .............. g E .uy... ..g. l Clflictb O

1':83 X0*13 ikGH 38]I lid 3 mt;31It4 uw-T-

.--S _w q e

6.0 Croup VI Events Group VI events involve reactivity and power distribution anomalies associated with control rod malfunctions. 6.1 Rod Withdrawal at Power (TuPIC X V-13 ) The inadvertent withdrawal of a control rod because of operator error or rod controller malfunction causes an increase in core power level ~~ ~ and heat flux. Severe local peaking can also occur. Protection is afforded by a rod block from the average power range monitoring (APRM) system of the reactor protection system. The ARPM system uses signals from the local power range priorities (LPRM's) to measure core power. When increasing power is decceted, rod block signals are generated. The Exxon analysis (reference 14) of a single rod withdrawal was per-formed with the following assumptions: 1. No xenon or samarium ^present 2. Peak core reactivity 3. Control rod pattern which maximi:cs the reactivity insertion. 4. transient rod is fully inserted, adjacent rods withdrawn The APRM rod block terminates the rod withdrawal at a set point of 106% power. No MCPR limits are reached for this event. This event determines the steady-state MCPR operating limits since the ACPR is greatest for this event. By passing of APRM channels and failures of LPRM detectors t I l e ,-w-

cro considered in this. event. The_ combination of bypassed and/or z . failed detectors is limited to those permitted by the technical specifications. . The following steacy state values are estaclisned in orcer'to provice for ~ ' ~ i protection of the minimum transient MCPR of 1.34 at the rod block. 't nica SfE' APPli System Required MCPR Steady-State sfiaci 9?ct Status At Rod Block a MCPR MCPR _Mc'c 1. No channels ~or LPRM 1.34 0.16 1.50 l-f1 i inputs bypassed. 2. A-C or B-D level channel 1.34 0.21 1.55 1.52 bypassed or 0 level LPRi input bypassed. 3. B-0 level channel by-1.34 0.26 1.60 l L'{ i passed and C level LPR4 bypassed. A rod block would be initiated either by the neutron monitoring system (NMS) or by the rod worth minimiser. The NMS, through the reactor protection system, inhibits rod withdrawal when too high a flux is detected by the intermediate range monitors. The rod worth minimizer ~ initiates a rod block if an out-of-sequence rod is selecte,d, or if a rod is withdrawn one notch beyond the pattern position. 1 The RWM syste= is basically provided to minimize the consequences of a rod drop accident. However, it also serves to protect against rod i withdrawal errors during startup. The RWM is required for up to 10" l power. i This equipment ef fectively prevents a continuous withdrawai which could l be ll=iting. Analysis of this event, therefore, is considered only in the FSAR. I l l e-e -r-e .-r-v

l..

1 &C AWAL YJr.$ of -71/[ Sc0 idtTHOA4WA G A1' I'o'di~4 EvCA T /$ GQ ulva L@'i'

- Z~2 C.URKC^l T*

HCC CdiTEd/A. Mrs 13 /3A JEO e4' ~~7~M t* FoLLO t.d id y: btMILAA. Codd,c/2 /A 77 /s A SSuns* 71043 Alc~ AAOC L J/1"M f [ 3 A E C 7' 'T5 / Hi TIA u. Co N O t ~n od 5 ~ . As isc.cerraatc catcue s-rionse moscal s<:- useo 114 7DE /7 HALY Q - r ter SAsa. 'i'Er i4ccep3NCE CKIT*e~icm foi?. -nw Gva% 7~ ACE r s e = %0 4 0 t a 4 o G 4 g + 9 s a 4 i. t i m*

6.3 Rod Droo-(XV-13): A rod drop accident' occurs when a. red is removed from the core at a more rapid rate'than can be achieved using t e drive =cchanisms. A fully h . inserted rod is a*ssumed to drop out af ter becoming disconnected from its' drive. The= rapid reactivity insertion cause a flux spike and rapid increase in energy deposition. Fuel damage could allow the release of fission products and have potential radiological consequences. 7 The adverse effects of a rod drop are limited by the-rod velocity-limiter and the rod worth minimizer. These devices ensure that the reactivity insertion due to the rod drop is as small as possible. The_Rh*M is used-l only below 10% power. 1 No automatic reactor trips occur in the rapid time frame of the event. Doppler feedback is assumed to terminate the event. 1 The worse case occurs during hot standby conditions, with the vacuum pump operating, since this provides a direct pathway for release of I radioactivity to the environment. Also, the rod worth is high for ) .. conditions. i j The rod drop event was analy:cd in the FSAR using General Electric j~ excursion analysis models. Fuel rods with enthalpies exceeding 170 . cal /gm were assumed to experience eventual cladding damage. This is i in agreement with the Standard Review Plant for BWR Rod Drop (15.4.9). I The analysis was performed at 1600 Mat. The radiological consequences we: later scaled to a power level of 1930 Sfwt. I i. I l l l-l b l - N ame _ _ _ _ _ _ - _ _-

n. ;- __

__-__-7__ -w _ + - _. _ - - =.....,. . ~ . ~,.

The acceptance criteria for this event as given in the SRP are that: (a) the reactivity excursion shall not result in radially averaged fuci rod enthalpy greater than 280 cal /gm at any axial location in any rod. 4 (b). the maximum pressure during the transient shall be less than the design limit. No fuel rods are predicted to have enthalpies greater than 220 cal /gm. For the hot s andby case, the increase in steam flow due to the energy release can be handled by the turbine bypass, so pressure does not approach the limit. Thus, the criteria are satisfied for this event. Due to the fast response to this event, no immediate operator actions can mitigate the consequences. lollowing the event, operator response is di-rected to recovering the rod; and maintaining the reactor in a safe ~ condition. 6.4 Radigl_igicp1 Consequences The consequences of a rod drop accident were calculated in reference 1 and updated to 1930 FM in reference 5. The consequences have been evaluated and the design of the plant has been found to assure that the recovery from the accident is sufficiently rapid and effective so limit the activity releases. The evaluation cf

~_ n. f radiological consequences.has been performed using an analytical' model based upon a conservative description of the plant response to the accidents. The calculated doses are presented in table 6./-1, r and are well within (taken to be less than or equal to 25%).the 10'CFR'Part 100 exposure guidelines. TABLE 6.4-1 i Calculated Doses for Rod Drop Accidents. I Peak Of f Site Doses (rem) 2 hr. Total ) Whole Body Dose 5.0 X 10-1 1.0 Thyroid Dose 2.2 X 10-0

8. 3 }; T' 1

1 1 ) e f a t 4 +.,, -, -,,.. p7 m -

7.0 GROUP VII EVENTS These postulated accidents result from a loss of coolant, in_ excess of mak 2p capacity, due to piping breaks in the reactor coolant pres-sure boundary. E7.1 Loss-of-Coolant Accident A loss-of coolant accident-is caused by a leak or rupture of lines containing primary coolant. A full spectrum of break sizes are con-sidered, up to the complete double-ended rupture of a reciruciation line. The loss of energy and mass from the system causes vessel de- ~ pressurization. Continue'd loss of coolant would lead to core uncovery, .and would prevent heat removal from the reactor. Protection is afforded by emergency core cooling systens, which are ~ designed to reflood the core-following a break. The reactor trips on low reactor water icvel or high drywell pressure. The course of the accident depends on the break size and location, whether offsite power is available, and the assumed single failure. ~ The analysis considers ECCS actuation with the diesel generators sup-plying the power. Far a small break, the feedwater system is used to maintain vessel level if offsite power is available. The isolation condenser are-used to remove decay heat from the system. The alternative is the automatic depressurization system (ADS), which reduces reactor pres-sure so that the low pressure systems can operate. e l 0* .,----,_,n- - - - - -,, -, -,,, - _W -~ r-

~ ..r The small break model assumes loss of of fsite and auxiliary power i coincident with the break..This results in coastdown of the re-circulation pumps, trip of the main feedwater pumps and closure of . the main steam isolation valves. The condenser bypass valves are assumed to allow steam relief for 1.71 seconds af ter loss--of power. The c osure of the main. steam line isolation valves results in an immediate reactor trip. The small break spectrum analyzed is f rom 0.1 ft to 1.0 ft2 The breaks were assumed to have occurred down-2 stream of the venturi with failure of the isola; ion condenser valve in the broken loop. Small break spectrum LOCA analyses using ENC's NJP-B'JR-E.M model have shown appreciable PCT reductions relative to previous analyses which demonstrated the maximum PCT conditions in the small break spectrum. f This condition arose due to the use of the Ellion pool film boiling correlation during the blowdown transient in the earlier mode'.. In l i the present model, the blowdown flow calculation results in a nucleate boiling interval during which much of the core stored energy is.re-moved which results in considerably reduced clad temperatures during I the subsequent core spray interval. Thus the highest PCT is reached I for the large break. I 1 For very small breaks ( < < 0.1 f t ), the CRD pumps may be used to supply 2 cooling flow. These pumps are automatically sequenced on the e=crgency i diesels. The analysis, however, does not take credit for this system since it is not safety grade. 1 [s m**+-. s..', n. ..._,m_, .m.~

..e l 2 In'the 1.0 ft blowdown, the intact loop emergency condenser remains r i on once initiated since the valv.a flow magnitude is less than the automatic shutoff criteria (flow 2 275. lb/sec) 35 seconds low water level signal. The automatic depressurization systems-(ADS) is not -activated: Core rated spray th >recched'88.6 seconds af ter break initiation. 2 For the smaller breaks (0.35 f t and 0.1 ft ), the emergency con-denser shuts off early in the transient. Because of the slcwer depressurization the ADS systes is activated after its 120 second i delayfromtheloI-lowlevelsignalineachofthesebreaks. For I 2 2 the 0.35 ft and 0.10 ft small breaks wherein failure of the emer-gency condenser valve on the broken loop was assumed, the time of 1 l. core rated spray was 186.7 and 392.3 seconds, respectively. I e 1 5 i The assumption of the ADS valve failure rather than an emergency condenser valve failure has the effect of slowing the rate of de-2 pressurization late in the transient. For the 0.35 ft break the time of core spray initiation was delayed 5.2 seccads and the time of core ? rated spray was delayed 9.3 seconds frou 186.3 secons to 197.0 seconds after break initiation. Heatup calculations have been performed for each of the four small breaks discussed above. ENC 8 X 8 reload fuel at 7.0 G'JD/FE11 burn-up with octant symmetry was used in these calculations. The resulta indi-l cate that the time of hot plane uncovery to a large extent governs the e l- __s

j'... final FCT. The assumptio'ri of ADS fail ~ure rather than emergency con- ~ 2 denser failurc for the 0.35'ft break was found to result in a PCT ~ reduction of 16.0 degrecs. For larger breaks, depressurization through'the break drops pressure below'the low pressure system shutoff pressure so that core spray cooling can commence. ' ADS is not required for depressurization. Heat is removed from the drywell by the containment spray cooling system. ' Double-ended guillotine (DEG) becaks with discharge coef ficients (Co) of 0.4, 0.6 and 1.0 were considered as well as split breaks with 2 break areas of 1.0, 2.5, 4.0 and 6.292 ft The worst single failure was the los.s of one emergency condenser. The operable caergency con-denser was connected to' the intact recirculation loop recirculation-y_ loop based on sensitivity calculations showing this to be the worst location. The worst.large break (limiting break) was determined to be the complete l double-ended guillotine break (CD = 0.4) of a recirculation pump discharge I line downstream of the venturi. The worst single failure is loss of one condenser. A steam line break analysis was performed to confirm the assumption that recirculation line' breaks result in the most. severe LOCA PCT's, i The break are assumed was that for a full steam line pipe diameter { (2.54 ft2 ). A PCT of 682 F was calculated for the steam line break as opposed to a PCT of 2200 F for a recirculation line DEG with similar initial operating conditions. s 7-es w ,. l

1 l Two additional assumed breaks were calculated using the approved NJP-BWR ECC.c Evaluation Model, a core spray line break and a feedwater line break sn reference-24.. In the core spray line break, the spray line, the spray header The ring, and the spray.no: les were modeled in the RELAP4-EM calculation. break was assumed to be an open-ended break with a Discharge Coefficient (CD) 2 The small break model was used in the analysis (break area.196 ' f t ), of 1.0. The analysis showed that the reactor depressurized. rapidly until the system, pressure reached the pressure equal to the saturation condtions of the lower plenum.- After which the system pressure remained n~carly constant until the Automatic Depressuri -ion System (AD') valves opened,' permitting the system S 3 to again depressurize. Core rated coray, 468 lbm/sec, was calculated t o occur at 527 seconds. During the course of the blowdown transient the core remained covered with water. The peak clad temperature calculated for the core spray ' ~ ~~ - ~ line break is 13350F. For the feeJwater line break, the feedwater line, the feedwater sparger ring, I and the sparger orifices, were modeled in the RELAP4-EM calculation. The break was' assumed to be a guillotine break with a discharge coef ficient of 1.0. The ) guillotine break permitted flow out two of the four lines feeding the feedwater The feedwater sparger is located in 'the downcomer of the reactor. i sparger. 2 I The break flow area was (.998 f t ) hence, the small break model was used to analyze l the transient. This transient behaved similarly to that for the core spray line break. The reacotr depressurized rapidly to the lower plenum saturation pressure. I The calculated pressure was then nearly constant until the ADS came on. The tran-j I sient calculation was terminated at this time since the core was still covered with ^ water and the trans'icnt fluid conditions are nearly identical to the core spray line break. The subsequent depressurization rate is controlled by the ADS flows, ^ ~ ~ ~ ~ e> v

i l l f therefore, the core was not expected to uncover as shown earlier by the core l spray line analysis, and a peak clad temperature similar to the 1335 F cal-f I from this transient. culated for the core spray line break will result The core spray line break and feedwater line break are clearly not limiting for Oyster Creek. 4 = 6 m 4 e =

== M -"m_

- - - N

.g g, i

9 v ) 7.2-Radiological consequences The censequences for this event were evaluated in the FSAR and updated for operation at 1930 is in reference 5. -The consequences were re-evaluated using the assumptions of safety guide 1.3 in reference 29. This calculation was also done for full power operation at 1930 3M. f u ce(cresce Ei .The LOCA evaluation using these assumptions results in the off-site j doses tabulated below: Site Boundary (2 hcurs) Thyroid 145.0 Rem l Whole Body 9.5 Rem l t Low Population Distance . Thyroid 117.0.<m. (30 days) Whole Body 4.5 Rem These doses are well below the guidelines of 10 CFR 100 even when the conservative assumptions of the Safety Guide are used. e e asse m. w } s 4 l l l i e w, y g._ _7

- 1 1,*. 8.0" GROUP' VIII EVENTS .1 These eveats are infrequent occurrences that lead to possible radioactive releases from fuel damaged by dropping a heavy load or through fuel handling. 8.1 Radioligical Consequences of a Fuel-handling Accident (TOPIC XV-20). The refueling accident analyzed in the operating license ap;11 cation state occurs when a fuel bundle is accidently dropped onto the top of the core during fuel handling operations. This is not the accident .re. quired by Requlatory Guide 1.25, but the resultant fuel failure.and t f radioactive fission product release is at least as extensive. A com-I ,j-parison of the safety analysis for this accident ~and Regulatory Guide 4' 1.25 is provided reference 32. f There are a number of area of non-compliance with the Regulatory Guide,

i but the =cthods and numerical values used are well documented and I

justified in the General Electric Topical Report, APED-5756; " Analytical i ~ Methods for Evaluating the Radiological Aspects of the General Electric i Boiling Water Reactor". l 1 i The potential dcscs to off-site persons corresponding to the fission l product release for the fuel handling accident are well below'the j i 10CF 100 guideline limits. Additional evaluation, with much highre i ) fission product release levels, has been performed (see the response to Question 1-13 in reference 33) and the resulting dosages were still i below the limits of 10CFR 100. k i -v 3 m..

The analysis of t'he refueling accident involves the mechanical damage ca2 sed'by a fuel bundle falling-back onto the top of the core while it is being removed, and the subsequent release of radioactive fission products. The severity of the consequences depends on the= fission product inventory-in the fuel-and various-factors af fecting the amount and kind of releases to the atmosphere. .There will be no change in the total quantity of fission products ~ ~ due to the change in the reload fuel design since it will be operating at no higher power. level, but there will be slight changes in the relative amount of different-constituents because of the presence of gadolinium in the Type VB bundles instead of boron in the curtains. The effects of these offferences will be small and undetectable when the various reduction factors are applied to determine off-site doses. The most significant difference introduced by the Type VB fuel design is a substantial reduction in'the quantity of gaseous fission products in the rod gas spaces due to the lower fuel operating g temperature. Given the sample bundle and exposure history, the fission gas inventory potentially availabic for release in the 8XS bundle gas spaces will be approximately 40 percent of the inventory in the 7X7 bundle gas spaces. This being the case, the previous analysis of this accidentg=co: r c m.. G ;. ; w. a d &T0 ^ A" -. -........ = 054 con-1 servatively applies t'o the 8XS fuel. e i ._-*M"G.D -.mA>w41D---_-, -._N N _. --L ,a e-3.v

1,'.. i -9.0 CROUP IX EVENTS These moderate frequency events are essentially small break 'LOCA's. However, duc-to the postulated frequency of occurrence, they should satisfy ' the f uel clad integrity criteria. 9.1 INADVERTENT OPENINd 0F A B'A SAFETY / RELIEF' VALVE IT c P 1 C X V-18) I -The inadvertent opening of a safety or relief valve results in a . reactor coolant inventory decrease and a decrease in reactor coolant-Neutron flux decreases due to additional void formation. systen pressure. The pressure regulator senses the pressure decrease and partially closes the turbine control valves. No trip occurs, and conditions stabilize at a power level near the initial power. The feedwater system is used to makeup the continuing loss of inventory. Opening of the turbine bypass valve is -less severe since the capacity is 4 a less and the pressure regulator can respond faster as the turbine pressure drops. If the pressure regualtor fails to respond, the increased steam flow would cause a decrease-in steam pressure and close the MSlVs, such as a discussed in section 1.3. If a relief bypass valve sticks open, the continued depressurization could I require a reactor scram. If possible, the operator should reset the pres-sure regulator downward so that as much steam as possible is diverted through the. steam line to the =ain condenser (MSlV open and of fsite power), rather I than through the relief valve to the suppression pcol. b l O -n--~~er -, a 2 -s-, n - m.- -,--.4,:- -em-e

The licensee performed analyses of the opening of a relief valve in Section B.IV-2.3 and opening of a bypass valve in Section B.XI-3.4 of Reference 5. The safety valve opening event would be very 'imilar s to a relief valve opening since their capacities are about the ume. <k /D5a~u ' cA. f au. STa:Sa um r 7~.5 .A/:d <?NM *"6 ~ e em Om a b 4 e e s 9 e or ,up, _ _, - _ am -n.~ - ~.-.--- -ex W

?- I f. ONE RELIEF.VRLV E I 150. RCC10 CPENING 1 930 i NEUTRON FLUX i j 2 SURFRCE HERT 7 LUX i-RECIRCULATIOi! FLOW ( -L s jig _ gf EE0;ED;iil tVESSEL,, S a s ]r l ~ 7 C < uJ 100. 3 3 3 I 3 3 i-1 g-OYSTER CREEK NUCLEAR POWER PL ANT k ( UNIT N0.1. F ACILITY DESCRIP Tt0H (L AND S AF ETY AN ALY515 R EPORT 1 Inadvertent Opening of Ilclici e Valve - Systern llesimse - }--- 1930 : cat. I> tot 1 Reviced Dece:nber 19'to ,7 i FIGUltE IV-2-Ca go 50. cc L1A (L. ~ o smo+ 5 l'-. ~ ) O. 8. 6. 4. 2 i O. TIME (SEC.1 =

a 1.- o y ONE RELIEF VALV E 4 ~ 100. ACCIO OPENING 1 330 VESSEL PRESS THRNGE. PSI i RCT. SKIRT SUE iERSION IN. 2 OYSTER CREEK HUCL err PorER PLANT UNIT NO.1 F ACILIT Y DE SCRIP 180N AND S AF E TY AN AL Y$li RE PORT Inadvertent Opening'of Iteltet i Valve - Svstein Response - 1930 lo:t - i ioi 2 50. Reviced recert.Ldr 1970 FIGUlt t: IV-2-Cl3 d -2 2 2 2 - j 7 O. ~- 3 1 1 1 s e< 1 6 4 ~ I'. -50 2 4. 6. 8. 0. TIME (SEC.)

. q, .?, 10.0 CROLT X EVE';TS cGroup X events have a moderate frequency of occurring and lead to an ^ increase in pri:$ary coolant inventory. These events-could cause an Jncrease in pressure and power. 10.1. Inadvertent Operation of ECCS Increasing Core Inventorv (ToPtc. X6 I The high pressure emergency cooling systems for Oyster Creek are isolation condensers, which rely on natural circulation, and the feed-water system. The low' pressure core spray system cannot deliver flow to'the vessel until pressure drops to approximately 350 psig. Thus, this event is not analyzed. The increase in feedwater flow transient, which increases coolant inventory, is considered in section 1.2. O O b e e e t l e G l s -.n.

r-i stlSLOAQld(r of f'UE L A.sS6M ALIES e,,, ,] 'l~M G MMP ' X l LV EAJT*.5 ~ t H f0 L n't e Id THG Co KE. 'g/g e c riLTfD GRROKS CouLD LEA D 'TD Po wer 11,$ TGL MTSLOADING G~e< N'0 TAI 30 *IiOA ANOMALIES Ado Gx cCaipod & f~uGL. LIMIT. n io 'The inadvertent loading of a' fuel assembly in an improper position could increase the fuel assembly power because of the' difference in .the water gap and the exposure / enrichment mismatch. There are two types of inadvertent loadings considered in this analysis; mislocated and misoriented. d'misoriented. fuel assembly is incorrectly rotated' 90 or'180 within the fuel cell. The mislocated fuel assembly.is-loaded-into an incorrect location within the core Both fuel loading errors result from multiple ~ operator errors. To insure a fuel assembly is properly positioned in the re'ctor core,. l a I-the trained fuel handlers performs the following visual checks: 1 i

1) The channel fastener is located at one corner of each fuel assembly

~ ~ adjacent to the center of the control rod. - ~ ~~ ~

2) The identification boss on the fuel assembly bail points towards the adjac'ent control rod.

a

3) The channel spacing buttons are adjacent to the control rod travel area.

~ f '~ i

4) The fuel assembly serial number, located on the bail, are all readable f rom the direction of the center of the fuel cell.

s

5) There is cell-to-cell replication.

f I In addition to the five visual checks, the location of each fuel assembly 1 in the reactor core is visually verified in accordance with an approved procedure after the reactor core is loaded. The indeper. dent verification identifies each fuel assembly unique serial number and. compares its location to the loading pattern. e l -=- m,,,,,-

. g.. .?^ A 'loadingierror could ' also be ~ detectable by an inventory of _ the ' discharged fuel assemblies. The effects of a mislocated: fuel assembly at the Oyster Creek Nuclear Power Station has never been analyzed. However, General Electric has performed numerous fuel assembly misloading analyses for generic and-plant cycle specific BWRs. The results of these analyses have never resulted in the postulated violation of'the safety limit critical power ratio. - As a result of-these so i - , analyses, General Electric informed the NRC (Ref. )j ) that they will discontinue performing plant-cycle specific mislocated bundle l i t analysis. . The misoriented fuel assembly accident has been analyzed and documented for the Oyster Creek Nuclear Power Station on numerous occasions. The initial analysis appeared in the FSAR, Ref.,1_,, for the original 7X7 fuel. This worse case analysis conclused that a fuci bundle misoriented 180" would result in an increased bundle power of 29%. This increase in bundle power does not' result in exceeding MCHFR limits. J }l FO an updated analysis concluded that a misoriented bundle In Ref. f + vould result in an increase bundle power of 21% which remained within MCHFR limits. In Ref. /0 it was shown that a misoriented Exxon 8XS VB fuel assembly would experience a 17" power increase in the worst case analysis and is conservatively bounded by previous analysis. I t I i k l 6- % W WP - 66_ _ _ y y.-- .,y y .,.--y ,y. -r.

P. O., REFERENCES- '1. '. Facility Description and Safety Analysis Report, Final, Amendment. J 3 for the Oyster Crcok Nuclear Power Plant Unit 1, January 1967. 2. Amendment 14 to the Preliminary Safety Analysis Report, December 19, 1967. 3. Amendment 55 to PSAR, May 5, 1970. 4. Amendment 63 to P2AR, September 17, 1970. 5. Amendment 65, December 1970. ' 6. XN-74-6, Revision 1, September 1974.

7.

T. S. Change Nos. 33, 35, 36, January - March 1975. 8. XN-74-38, Revision 1, September 1974 (Note: B, 9 and 10 transmitted as part of Amendment 76, January 31, 1975). I' j 9. X:(-74-41, Revision 2, January 1975.

10.. Amendment 76 (Supp. 1), March 25,1975 l

11. XN-74-43, Revision 2, January 1975. A 12. Amendment 76 (Supp. 3), May 1975. 13. Answers to NRC Questions, May 8, 1975. ~ 14. Amend =cnt 76 (Supp. 4), October 20, 1975.

15. Additional Information Re New Proposed MCPR Limits, April 15, 1976.

I t M b f e gu. %. em -- eDe gdun enme e.- M v i,w., y -,

gg

t- .m -; 2 _ s 6 >; 16. License Amendment 9, Change No. 25 to Tech. Specs.,'May-24 1975. ~ 17. LOCA Analysis Model' Documentation, March 25,-1975. 18.. License Amendment 8, May.24, 1979 19. Amendment 15 to POL, February.24, 1976.

20. -

LOCA Analysis Re-evaluation, Change Request #40, December 23, 1975 21. Amendment 16, July 26, 1976.

22. - XN-75-55 (A) and Supplements 1 and 2, dated August, 1976 23.

Letter-to ENC, March 28,.1977 24. ri-NF-77-55 Revision 1, March 1978 25. Letter to NRC, answers to questions 2/5/76 26. Letter to NRC, transient of May' 2, 1979,'5/12/79 - 27. Letter to NRC, Transient ~ of May 2, 1979, 5/12/79 28. Letter to Mr. S. Norwicki (NRC), f rom G. R. Bond (CPUSC), May 17, 1979. 29. Amendment 68, March 6, 1972 30. Letter to NRC (T. A. Ippolito) from CE (R. Engel), Change in General t. t Electric Methods for Analysis of Mislocated Bundle Accident, dated ( November 14, 1980. r. l l e I..- ~

1 31. Oyster Creek Facility Change Request No. 4, dated January IS, 1973. 32. Amendment 68 (Supplement 6), November 1, 1973 33. Amendment 11, June 21, 1967 9 e o 3 er e e e ( 4 4 0 e_

r-- .Oi ef ' TABLE 1 T REACTOR PROTECTION SYSTEM (RPS) SETPOINTS SETPOINT PARAMETER 115.7% High Neutron Flux, High Reactor Pressure 1060 psig High Containment Pressure -d2'psig Low Reactor k'ater Level 11'5" abov,e top of active fuel a Low Condenser Vacuum J.23" Hg Main Steam Line High Radiation' 10 times background Scram Discharge Volume High Level 37 gallons 1 Loss of AC Power to RPS Closure of MSIV's 10% closure i Turbine Trip 10% stop valve closure Load Rejection Loss of oil pressure from turbine acceleration relay Manual Rod'Bloci 106% Scram delays: 0.2 see Scram Mechanism 0.3-0.9. Isolation Valve Closure (depending en closure time) e I

W^ ~ ~ -yp .f' TABLE 2 ENGINEERED SAFETY FEATUPIS Parameter-Setpoint Systems Started Low low reactor water 7'2" above top Core Spray level of active fuel Containment Spray Automatic Depressurization System Isolation Condenser (if signal persists ,for 3 seconds) High Cc.tainment-2 psig Core Spray Pressure Containment Spray in Conjunction with . Low Low Reactor Water Level Automatic Depressurization System in con- ' junction with low-low-low' Reactor Water level Reactor High 1060 psig~ ' Isolation Condenser (l'f signal persists Pressure for 3 seconds) Low Reactor Pressure J285psig Core Spray Permissive ( j l j u.--~.--.--- - -_=. a.- . r.- __ m

G, TABLE 3 CONTAINFENT ISOLATION Systems Iso'ated Parameter Setpoint l liigh Steam Flow 4_ 20 psig A P Isolation Condensers liigh Condensate Return Flow f,27" AP Main Steam Line Drain 120% liigh Steam Line flow Main Steam j Reactor Building Ventilation 4 liigh drywell pressure $2psig Sumps, Vent Purges Traversing In-core Probes. Reactor low-low water level 7'2" above top Main Steam of active fuel Main Steam Line Drain j Cleanup System s Cleanup Auxiliar Pump System Shutdown System Reactor Buidling Ventilation I. l Steam Line 'digh Radiation 10 times background i Main Steam t Main Steam Line Drain 4 b 4 i

l = (. i 1 I i TABLE 3 (continued) ((: i } CONTAI; DENT IS01ATION 8 Ii t. Parameter Setpoint Systems Isolated ,g High Temperature in Steam Tunnel 500F above ambient Main Steam Main Steam Line Drain i l i 4 a ~ i High Reactor Pressure >120 psig Shutdown Systems. 1 f. 1tigh Reactor Building Radiation 'f,17 mr/hr . Reactor Building Ventilation 1 l l-liigh Radiation Reactor Building 6.100 mr/hr Operating Floor Low Steam Pressure .h 825 PSIG Main Steam 4 '4 i l i i t I i ~ s I J. i e g

'a s .=m TABLE 4 ANALYSIS ASSUMPTIO"S Event Assumptions Decrease in Feedwater Temperature 135 F feedwater temperature, 100% power . Increase in feedwater flow 5's:-power, 42% flow, 110% feedwater flow 110% of rated steam flow 96.4% power, 45.7%' flow Increase in steam flow and hot standby 0 Startup of an inactive loop 10G! power, 10G! flow, 100 F water in isolated pump Flow controller malfunction- _ 53% power, 42% flow, 10%/sec change increasing flow Main steam line isolation 100% power, 3 sec. closure time,_no credit for valve closure isolation condenser Loss of load 100% power , Turbine trip 100% power failure of bypass failure of relief valves failur'e of isolation condenser (scram for safety valve sizing transient initiated by valve position switched at 90% open) Loss of condenser vacuum 100% power Turbine pressure regulator failure 100% power mf Loss of feedwater flow complete instantaneous loss of feed '"' - _ 100% power; Initial water g, r--- icvel just above reactor alarm set point and abount ene foot below normal operating level. ~ Steam line break inside drywell see LOCA Steam line break outside drywell 100% power 3 sec. isolation valve closure time Rod Withdrawal No xenon or samrium peak core reactivity maxi =um rod worth transient rod initially fully inserted, surrounding rods withdrawn ,,s -=-h4--}5 y f .3 vu- -m..pm .se- = we 4 --e*-..s-- . hap, -. " = " O

~ '- ' ] ..,, _ 4 j i .) i 1 -Table 4 (Continued)- Event Assumptions 1 Rod drop-Hot standby' maxi =um rod worth i 1 1 ~Loze-of-coolant accident' ' 102% power, worst single' failure of ~ ECCS, loss of offsite power. Inidvertent opening of'a relief valve 100% power Inadvertent operation of ECCS in-see increase in feedwater flew creasing core. inventory Loss of auxiliary power 100% power Loss of forced flow 100% power O ' m a e- &mm 4 i 1 m - 9 4 e 4 e 'e 4 1 r-. .m -s'.U- '-'?

~ ~ - '*

'}}

ML20039A385

Contents

Text

Dear Mr. Fell:

Response

Response

Response

Response

=

3.1 DESCRIPTION

==

Navigation menu

ML20039A385

Text

Dear Mr. Fell:

Response

Response

Response

Response

=

3.1 DESCRIPTION

==

Navigation menu

Search