ML19326C000

Amend to License DPR-51,providing for Reactor Vessel Matl Surveillance Program

ML19326C000
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/01/1977
From:	Goller K Office of Nuclear Reactor Regulation
To:
Shared Package
ML19326B999	List: ML19326B998 ML19326C000 ML19326C005 ML19326C012
References
NUDOCS 8004180693
Download: ML19326C000 (11)

Text

_ _ - _ _ _ - _ _

A w g e..-

g g

.g g l

agggggggg%g.wmp ywwgggqQ p w;y.4.ue g.

&,x tg..x.g.,k.-w&;G &mw _ w,.qymgy gqqq%

c,;#

y p:

i

.pz y

. -d :

M M:M W,; M Q:d$7 M,

gass;M.kMM%$MMMM.N...' :eggewgmp NfMN,.5.

+

WG&

l jd 3h 1

7 p.

Q.

r. - 4..,.

a.g.n

%.,,.m,% g {,

..Qe ARKANSAS POWER'A LIGHT COMPANY g W M.n g

l

. s.

1

.nv..,,.,y,__

s.

e,~

s

__2+-

a,a. x, x.. ; - g,,:s..yn., 1,, m.,._,.. -

--, c z.

l.

.mgg. e.t 3,;..gp _

j kne

...~

a.

~

.,.n z

u a

y q00CKET NO. 50-313 y h,,Pe;3g;p;w 3-ggpggggggggggggg@afy@;$rfph.mg.r Jf Mb.l i

~ mWU

, y.1 Sp. ARW5As. NUCLEAR ONE, UNIT N0.11 Tf M

A

1. 9 -

I 74 l

• g.

w e y m.g. e :p x. h h lEllBNi F TF FRETUTY 0PERATI K LI'CEMSE yd;h 2

pf,

m gun.4g w-o$ -

I w M. W.p + 6r.y -

w

%_,a NM/. f @

9

.,7 g.

e wp myv r. 2 'ggncom%jk.;;-M..'w g L,i en. ~,Ms.

w:-

W cr ah W

9:

4 1 w;e r w w %w. q, u.

e..

k g f y,> j R

• M o<i k @ m f., s x..M & M.Q y Q M y QQg v

c se W fM7cb[.a.7; i

m.

x--

..e www.mn r.m

.. n.,,

ex.

c 1.

The Nuclear Regulatory Commission (the Connission) has found that:

' y %

p'-

~

A.

The application for amendment by Arkansas Power & Light Company I

(thelicensee)datedAugust 17, 1976,. as. supplemented by letters 4

dated DecemberJ20 and122L1976,Tand JanuaFy 13? 1.977",'comptiesG,645d[m

'with the standahls and requireme'nts of the" Atomic Energy Act of f7T' i

1954, as amended.(the Act)', 'and the C'annissio~ 's rules and v NWar n

l regulations set 'forth;in 10 CFR C' apter;I;. ; Q" M645AW NyM _

n l

~w

. w-a

.m.. _#

... ym,,

,m.

~

+,.

wa g.

E' w t.

.. Js...

3 B., The facility.will operate in conformiti with ~ the applidation,F

%c. w.

v-

+

s-a Acte and ;the rules-andiregulationsueW %NS -

, fthe Caunissioni.J.N;q mypi"the previsions of l

e M

i 33;;Q %mmy s.JO,.C.%There,,is ;ggwC(e;)a;G;sq:.we@egm mMlttiest L

QMargb 3@;%p&b ~ &,y'

• Q*

mn,.

1 c.

ssuransaa;(y);dsinise*&mismihpT

~

U rgegsenal thetrthexacAty s

MN9@$4.nW3sa.h. ti,',s.ti. tiieT pi@sili,iM,fi.)f bui.t9eMtFadt. i. n$~

fehthfs?'amindmase si eemistle F

e M'~". M,, w cens4.ance withittet.Camrission'~s regulations wNNg.J +.g' t

t 1

a..

2 s ~

conducted iny.

.. Q W M : l Q & f & h M b Q h Q t!.% " W; +t h & ff% M.

u e.

D. ' The issuance ~ef. this amendment will not be sinimical toithe commen % M

.. defense'anddeburity or to. the[o "lthiand safety o V

2 e

hea ym ag.

g i

w%

w

,,4 -} c..O,:.Q) Q&7.y9WQ. Mff, w:Qf.QQ QL9?ll.QtgX l

. z.

.. m

._m E.The issuaneeiefathis. ~ f, ;jamendment?is in accordanceJwith 10"CfRT *JM?.%g 3

. i[Treguiraiien. Shef C$pMsstenfs;fe$ilati Pdt 51Lef 2.% - W B G N ts" hive 'Beeni'sati'sfiedE OM%W 1 4

4 ggggg.

2; Accordingly,; the31Jiemoe.fifsramendedihy?lchani es"tCtheiTechat;ea f

m

amendment, 'and Paragraph Ec.(2)'eflacility OperatiEg Licensef ' f"MO Specifications asaniHeated..in the 'attichment'to'.this 'ltcuse4W Mb M

No. DPR-51.1s hefebylamended

to. read.as follow 5: ',.;m!

~

1. QMMO e

.m 2,

v.

4 _

m

-sg, a

%y%%...

7,

w..

-: 3.g.- o.

f

,g e

g ; g.7 -

2.-

+

r Y __ '

'*7 M, w.

~..

...r.

...~.

.. s um m.

s- - -

L

^orrica b

"~#

[ A 1, N*

j.

m

.Q,ym

[<_

.x yy a

2-Foran AEC 318 G v) AICM 0240 W u. s. eavammasamteminvine orricus son.ose-see

^ f..s.. g w?":;,

• s

+ n da.cr.

~

1 s~

.J:

W %, O %.xy/.m$ $$,$n-y@{ $u ]j%u @.m. % $ $ g,[m mg@yg l Q %m.47

.h[hh.ga..u..;.Ma&_ a

. u%.. b w

u w

h.. a.

h h: h.w w~ hf$h.h.hf)p,gf

.m,w w n

- M wwpAw:s--y w'N -k:\\p.:m, m.~+f m.m 1, n

n r~

%, w@.. p -

y

1. ~y..

9 4p ?(;g%7 ar w2'2w. f.. "_. Qgg

%w M ~ W;W.%w Q y.i8%p,&y%.,;.f k #e w.m D.

4 h]h

,6~ s W r-i

' m ;:.y ;.,;v

, ;s x ;

b.,,A f _ tg ^.. ' w' Q, a,y y f;,e..g *g Qs 4m,#Mpe,ge,t_. ;o.i;s 3 _,.g @ gy m'.MMf4

. ~:,-

s n

" w ) ;M, *k, a

..a

, ~. w $ *e.n,e4 N, +y.c&, m.h

...y

, f. y, f.

M

. M, i 3

WWA m?;QMO.@i,M,. MMnMs@, q -.

,4

~

p~.w,w.. w,,.o, m;p%.a ~ m...,w,. y.s..w.a. d 3

W_

y,...w

- ~.

, c.w 9x.

xw.

g..

c. w v,

- n.

t q

, o-i m3.

3

-s

... 3 y.x.. v g

, e my

,.msnm.

M f;@v-v a#f4;gpymg:;gg

. ggg ymp

~

s2.

~a&-. > 4.a;n

. (2) Technical Specifications,

W6 W

p.W. c A M

%W.' fiKQQ WG pf&n u.c.s.F.4.M%

'QQ.y&?mj^

w

. q)@y z us, BMWl R

I The Technical Specifications contained in Appendices W V ,

A and B. as revised through Amendment No.

, are hereby A

l incorporated in this. license. The licensee shall

' MMMM l

~ operate the facil.ity in accordance with the Technical utN#M

$Wh m y, a - a. ~,? n@ n i4 % s f &.. g, &

fR*+

Spectf.i.ca= tion's. w ws. &. sg n

x n.w.w g...~ ~.n.x w.,f dtsews%, u

~,. N n, m+'

n mm.

n

+,

a r.

p

w. w 4

2 e.as^of thei datere.-

3. ~Thisclic. cense

- ohm,': amendment 31su,effectiv.

s J g 4.yc p Mm ww y

/ [pe 4d w. c w m : m a p. issuance. e.pw g a

-w mW m.

g ORggWC1. EAR RE84tATORWC000GS870It*

e f ??Y,. M, s.

. e.-

m.

v.

m y 4.h,m p ~bhgmr m y a

% h &m.m m:{ h.h.

Y

.f.^

x 9 &nn&..

f n.

.Q n%!..w.

e.w. v y.y.eEn wMrW M%QQ,wpW A W N % & M 2 W.',h M M W% Q

'l

+w

- Kar1~ R. Goller, Assistant Director

.P+

N for Operating Reactors

% t ' M:~

i Division of Operating Reactors

Attachment:

~'

Changes to ~the Tech' nidal i s 'N4%h WT8% N M kk$h b Specifications ^

MMe WW WNM *4 MWp?.F*MW x m.,

c

.-c % y,. '_

l e n # ' w, a

_n s

.n s

g y p y.,,

Date of Issuance:

vb-we

";m~

-' - v.; A. ' C.

.n s

a

. e. n %m.;

i g-ef.

j.3':..g}.

t;. 3

.f

~

z, - Q. ~ U.Q.

s

+

7

-un-

m-m. 4 f.x f..e s

. n w.

+

l 7 Qgh. $g j.M,g'~ 7+Q.ym.,nSk : Q&.

hs.y,-364..d.jfyAj dh,,$L..

~

M

<n n

~.f p a

.c

,. b, e, 3. g.,

. a.f. e. e* p

.,~g

. n., Y

+ v e.

. % p,

g, 4.,s., %.

4 yy

.A q

• jfQf:Qg_ Qf q tfQ, - e.

O.

g A

-I p

9 3.j - ;,

.m _

r?..

, ; ya,.., g,&,, s &, &n.r; Q:&fy&QQ(.&.

eg m.

w H

QQQQ

@u. ;p.m y a v '

Tr;p y. G.f*p;c%y@; ?(;>pym;w y!WFMjr&q h h Y s y DJ'y.hr V*g, 3

.offG%r%%%

,wineg@ j qno My,;&gV "

f 7'

' l.

y s x y.q p p,.... yn,,,;;&, s 7<%r j h.%m @ a rn g#.p.

r

.gg; gn,y M f,q

-c:

w p..+

w q-

69. &, 4..

m.%.c,w ' y M,g c.

,P s>% q.d. 9

g..- >)V{&,. w-

.m.

m u.-.

x n o.g g a,

w w

m-c

. }

y, 6

~ s.e,.

f

~s

.f-.

gr

+

.u

+ *.

., s - m,n-e

.-. n w<

s

.iJW

,,J..0.

- v w,. m.. %.

ea p,

. m m.-

c.-

~2-g<>

m.-

1 },.

i 11.

.,.iA.

l

[

.[, k, hd

[ w.[

8 )',I, q.-,

~-h k kU P4 Q

+

s g.

.M.. A$e a, :h$, $p ;['

4

.v a

~,..

d %..m...p f. ~, ^

, g &j 7

k. y.'g ; G Q*1 ~

Q 3., ~, u u p. - R W.g g y, m

o :n..... - t c G

.;. W s

s m.

n

+

- ? w.,.:y L.

. @<..C.. n M.

3 -

t r

., s. ;

J s-

-, e1 f

c.

.v. +: ~

' v 7

. ~,.,

t.*[n =.,y. N'.7.. g%

' b..

.,,y sp.f_

g s

=

'w.

4-,

Ngy ]d.ggj. g a,

. Q 3p*,I g [..,,, *d A"2 f.O

,9 i

n...q;. -. y.

', y + w * %v -

p

%,, 7-R,,*. %g.a..

x.

.u-

% ~,g... c., 3 3 x

y

.,y&7

,. :. c - - Q ~ a,.

[J.4

'. N' y.i c. M

w.

a c

..,$ y$5.,m..g & W[ k,,s, ; 4;

,! g

7 w;/

,'4,-f.

' 4 &,5 p.*

b r *

- n + %.. e. R rp '4 u.s & ' W h.f' '%..l t j J s. M. d M,L 5 S H W;.Q%W.2

\\

s,

..y Q

W m s' %y a.

i w W

,.1,v c. #.

.,.w. ; r., :.y 21

%,L 9.%,. w -y%.,, y,.,o.

1

~

m m

ng s

+

a

,,-, b.?,a. ~. m., s.yr ;,a g y; p.

s,y,

.,<.g' n

= -

,s ^

8

e. p.

~p g

-o s

, o m_-

'.,y ?

• q*# '^ '.

.~.-"%.,'.

.n3. - > o e

m t

r.

,yr-; _ u, 9 y

.. r,,

u,.

A - - ^ ', =

it

~,. :.

,'a-'#"

r

&+-

s ;>..v.

,y

,1,

. s

.'a. :s. _ s. k...

21

.v s,,

E 2

.., q'. '.

J " "*q g

g r *,

'g 4,

• ^

9 A,

.1 g, '

g 4.

' e. p 3

s

'Mr

+

Y,Y.

[L 6

[ 4.; v.

arrie s >

a s

~

. v.*'

u eunnus h C

n

'm -

s oars >

t. s s

,e,N

• l Forum ABCW5 (Rev. M)) AECM 0240

. - # u. s. sovannnesar eninvine orricas sen.sae-see

> e a r %" r

.g_.

'4--

9

@,p.-.

"Q.

,,, - d $is', 5].y. g

,,.%' e'. >

y

)'>

4' 7

.p.g.

A g

'f & M T W y@dp@ M E~? T.CM S,MRfW W M $ Epq

~.%

e gjnp%mW;Mewghwtw m y& W. W p;n.

1 M ~p: m, m ;. y.g %q W w h % g m p p; # $ y M f mm

. h M,6 W$8

' 96 M@ m %uplN5%.J,.,%,p$,+ M~

f.,. F9EMw QMWM Q M h.!m%.,.,+ Q.,a y p

%.Q f,n:QW.q2W.

d% y,4 3.:.s-Q u.

i

- n&~'m%.) 'ry 4,,Lg4',2@Ws,ed;M ATTAODIENT ID' tICl56SE 5MEMONENT N.

I*MASY'.:MM,.e% %m..M;

,, v., w**

., n

.,,_ f C '.,' w*e::..%?-

s ' e p:

-a % y *-

Kh

^*.

aQ r s ~J i ? Vg ; { M G}v,e :* *:4e. Q h

~*yps- -

8=':

j'

.s n.

e

ns

W;y%5W M y&dc&gggy

. M7:D,y,@ggg ;w.pfygg.ny v ve w,>+jtg3 ugny

.ywpma u

.xwy 9

y gg7,gg yg...gp x y

- g s{n. en*gw ;:

w; n a% s

,, E.,,u.,;>.T. m& u n ~/ h.

. n....

N>u,,

s, U..

R.

. S J 2,j

+gy.# ; r'ggMT NO 50--313 h,i u;M.Q W

:7 c

.n a, -

n

,N.,,.k.

A,n.~z w (w *w g e n.un.u;;e y ow. yw t ya.. a c ll.gJ;- W; c1%y{.m;,qs ~;

n p+ y ;. y;,,A c,

a c

2y i

n S.%

O

. v

-Th w y::.

a-

. w a rs y%..g@% yySf g QCMQTQN M.

?

+

Rep 4 ace the following pages of the Appendix A portion of the Technical

.1%

Specifications'with the attached' revised pages. ~ The changed areas on QYF?

the revised:pages:are,. identified by a marginal,line.

' g.,a Qn-@A.,s.M,

PG 4p m%cgy.

s

. 4 w, 1,g,%.

c m3 r. -

a-

~,

m.u 9. W "EMOVE n M s-e 3.cf..g w a t v u m y my.e i

e mg g yADO,4

-wr -

"%m~ M*M@,sc -%se.Q M.J:a%~,1 M h.f W4M W W$pww w.. m a m ~ 7 9 g*-

R v

ae.

l M

Y

f.

~%z, 3 w, m w w w q

.. I ~ < ~ n. m. n

~

n.

y-

i;, p. m. ;,, s v w W;fg$.,y~%m--

y,/r ; g;ay -< 20:;x.3..Abgpg.m,;ig T.g A;qwsq qag G;

9

,W::

w g

a.3

-*ff. ? A f g g y

y.,",M, p 0;n.w..

y,

. e. n w.t. em a

,5.yqqy,, s.,, g w q g g g g.y g.s...,

w m

20 x 77 77 4v 9A '

~

~,

77a

~ *

'('

77b I

.e

. e w,

g.g.g %, 9 3.g :p q,p g.-

- e.q ~

, ~,.syx, t

,r,.

m o e u...

.c y.s,,p.

y g..s...,n, m..n#;s W.. x..

2

..w m.v.

3-u..

u a

v m:

4,.

t

,. - $.~

,,r i v.

.:. n +.. r.:

s

+ <?

<

,. 3 p .1 u. .g . y mx u- ~... a n., c ,n _4 m. -r

• There were no changes on this page. "It'is included as a matter ' f.'

• lWT,p,

7-w- p q c,7 ,s. g g. e - ~ ~, e s < ~* 4 o convenience.,in : updati,ng t,h,e> Technical-Seecification,s,. mWh~MW.M'%v**m %.e ~,,. W 9 " w% s-i c m -. J.*~'<. a. ._ G, -->s\\ fe t- %'. s.'.,i fl&vy :'. c - S' ;J., .n - en. 's ~ ,? 7.* 1./ a ^- . y 3. .. > Q.. p;;'m & ( s> Q, p: f M g ia; g.: --& s 1 .; em-vzw.m : ~ m n,w;ty.*(~.[w*QQWy' y&;%.y - gpg-y.";'Qp,yQyy- ;;q.Q:

3. L.

.h. ey ir j j %.p n y

• e ~',;.Qg

. [. * #f ( +RfQ.: 57 l .)f,r <>-\\ .[- , p, : [ ? '..gh.@, M %yA, -

[r, %n.,s { $ E ' q / # e t

~ f f' - [Mh,d,y fg $,e.3h ~ h0 '~ ..a y.u9 _D4 > R. e. A6..s, . n, s, _. -s -c t' !...,-, x,3 :,.;, - - n, :m- _4 "n:E.. > g;:?.

~,s.e+N'f. y.w~r q <r p ' w.

n u I t ,~ ..., E '>s v.rt....y. '.N. X,W V x:;. ; - ?- 1 ,.. :n w n ~ .%, F Q.. ; *.k.,' h: ~,14C*.% 6,;f f . ?. .'r.- ='

s. -

.,..a e i .'*c'~ 1* !' d riwUt* 'lW in,W

• b ~,,

r . 4,h ~ E.c%-.m. p,',%,,., s..,,, c.-n--

n.. W'*,

+. : - < e r - w,.,p; y a.. _v. : s: c.,r,s, y .. 9. =s g y: p r,: m. l 1.. -er. t..r % : .-. g.: ~.w w3 v c -b w ,.* a. <.. s, c +'% ,~ r _s o. . ij, &W j (,l:f & ('?: &~,' Yv, '" g.n 4

& n 1,,, U 4,%g ?'

0 j W ' V.l h & his %. y< e T " ~ke c,j e 7,WM ^ ? .,.-6,e @ ; 1 *. 4.r. 4 ~ i s es,..:. ' " * %',g ~ 7..: g..y;;:: ,y y ,g-~,, _ f % % Q. 4~ ^g : e v m - r_yiqd;, ,. M %,.

1. g

} s% , 4[ s.* v: yg \\ -.y . O.,'.k.* + W -? 4 > ( .w. '--t. a b..h..hf.v .s h.2 ' Y _ k N, l s- - e f;*Ej rY y2< ., \\.f,3_ ;'.... w r - d .4 s Jg'.V ss : y-y(,,[ :. -* *f ',v*" s,;, ', l' ..h, .t ** ? ~ *;"- :_ a ' * ', '.,,. f y,'. *<.,.,f y' E' Q. &.eh,g l.W'.Q % nl

y. y g..y..;.n;3x pn

.yw_a _ v,,, v,:*-

• h k [;, *.

.l, + ,.g _ y u, .y

m. g w

- m%.. ~ .-r - q'u. / wgg,,. ;,* _g v3-s- y . +. .a ~ .j } x ?& f<., ,f;&ry], fA r.l w} f f l l

j

r.Rl ?,"_}-;p p&. k[.pm n,.jl t

i a .w....~'+.. t

• • • n ":, ad2:f' m..,s'.; Q*

• J,..,..i w.sf.t-;;

.,, i. n,.,7 y-f;.. " s. s - 1 . +. n s, ,..T. - % :>L

• i,9 _,

r#_ ,.,s+;

• .- 4

't. -- y, y. 6' . 'r 4 e. i $8 - CAg s 4 ^ tf p

• r

.,~ es x- .~ '.'a >:. :.yy. .i . a. ,,gA - %q r " <.e s'w ? "~ y ~> n. y y;, : .s -, u [,.,, s ! p:.-, $,5. s,.,, ..U- ~,M v'" S. ~.. P W'.f. m e. 3.- i .s p x.- s a .. y c h~[ I7 " *E- - Q ;,.,. orrec a w n ~,r+ f.-, s. -: *:v ~ .n . ^i - su mmaass > .c., r-- i- .~ ;

• _ ' ',ggg g.

..s 9,' 1 . x .c rw ' T f6 g7 2 Forms.AEC 315 (Rev. 9-53) AECM 0243 ' W.us s. oovannuant eni= time orricas nota.eas.see. < c. - de%f" ' - %;W.~ >. .s .n,1 3 + %...y. . m;g, o,. K;m ** ., as i y t 11a s e s_' 'A ~ reactor coolant pump or decay heat removal pump is required to be in opera-tion before the boron concentration is reduced by dilut ion with makeup water. Either pump will provide mixing which will prevent sudden positive'recctivity changes caused by dilute coolant reaching the reactor. One decay heat removal puup will circulate the equivalent of the reactor coolant system volume in one i half' hour or less. (1) 'The decay heat - removal system suct.cn piping is designed for 300 F thus,. the systea can remove decay heat when the reactor coolant system is below this temperature._(2,3) One pressurizer code safety valve is capabic of preventing overpressurization when the reactor-is.not' critical since its relieving capacity is greater than that required by the swn of the available heat sources which are pump energy, pressurizer' heaters, and.cactor decay heat. (1) Both pressuri:er code safety valves are. required. to be in service prior to criticality to ccnform to -the - system design relief capabilities. The code safety valves prevent overpres-sure for a rod withdrawal accident..(5) The pressurizer code safety valve lift- - set point shallibe set at 2500 psig + 1 percent allowance for error and each - valve shall be capable of-relieving 300,000 lb/h of saturated steam at a pressure not greater than 3 percent above t e set pressure. h The internals vent *.15es are provided to relieve the pressure generated by steaming in the core following a LOCA so that +ha core remains sufficiently covered. Inspection and manual actuation of the internals sent valves (1) ensure operability, (2) ensure that the valves are not open during normal operation, and (3) demonstrate that the valves begin to open and are fully open at the forces equivalent to the differential pressures assumed in the safety analycia. REFERENCES (1) FSAR, Tables 10 and 4-3 through 4-7. (2). FSAR,'Section 4.2.5.1 and 9.5.2.3. (3) FSAR, Section 4.'2.5.4. I (4) ~FSAR', Sec tion 4. 3.10.4 and 4. 2.4. ^ (5).FSAR, Section 4. 3.7. T 4 p 4 i Amendwint' No. 21 'l e ~s 4, r e. 4 c 3.1.2. Fressurization, Heatuo. end Dooldum Limitations gpecification 3.1.2.1-' Hydro Tests: - For thermal stccdy state system hydro tests the system may be pres-strised to the limits set forth in Specification 2.2 when there are . fuel assceblics in the core and to AGM3 Code Section III limits when no fuel assemblics are present provided: / a. Prior to initial criticality the reactor coolant system temp-erature is 10'PF or greater or 6 b. After initial criticality and prior to the accumulation of 1.7 x 10 thermal megawatt-days operation the reactor coolant system temperature is 215 F or greater.- 3.1.2.2 ceak Tests a. Leck tests may be conducted under the provisions of 3.1.2.1 abOVe or 6

b. After initial criticality and prior to the accumulation of 1.7 x 10 l

thermal megawatt-days operatiin the system may be tested to a pressure ofj 1150 psig provided that the system temperature is 175 F or greater. l 3.1.2.3 The reactor coolant pressure and the system heatup and cooldown rates s (with the exception of the pressurizer) shall be limited in accordance v.ith Figure 3.1.2-1 and Figure 3.1.2-2, and are as follows: i Heatup: A11ovable combinatior.s of pressure and temperature shall be to the right of cnd belov the limit line in Figure 3.1.2-1. The heatup rates shall not exceed those shown on Figure 3.1.2-1. Coollown: A11ovable ccabinatiene of pressure and tenperature for a specific cooldown'shall be to the left of and below the limit line in Figure 3.1.2-2. Cooldom rates shall not exceed those shown in Figure 3 1.2-2. 3.1.2.h-The seccndary side of the steam cenerator snall not be pressurised abovel200 psig if the tempercture 'of the steam generctor shell is below 100 F. 3.1.2.5 Tne pressurizer heatup and cooldoen ratc= shall not execed 100 F/hr. She'sproy shall not be uced if the temperature difference between ~ the pressurizer and the spray fluid is creater than h30 P. 10 Amendment No.. + / 6 3.1.2.6 Prf or to exceeding 1.7.x 10 thermal megawatt-days of operation, Figures 3.1.2-1 and 3.1.2-2 and Technical Specifications 3.1.2.1.b and 3.1.2.2 shall be updated for the next service period in accordance . with 10 CFR 50 Appendix G, Section V.B. The service period shall - be of sufficient duration to permit the scheduled evaluation of a portion of the surveillance data scheduled in accordance with Speciffcation 4.2.7. The highest predicted adjusted reference temperature of all the beltline region materials shall be used to determine the -adjusted ref erence temperature at the end of the service period. The basis for this prediction shall be submitted -for NRC staff review in accordance with Specification 3.1.2.7. 3.1.2.7 The updated proposed technical specifications referred to in 3.1.2.6 shall be submitted for NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR[ Part 50, Appendix G,Section V.C. BASES of cyclic loads due to system temperature and pressure changes. {pe effects All reactor coolant system components are designed to withstand These cyclic Amendment-No. 18a + loads are : introduced by unit load transients, reactor trips, and unit heatup and.cooldown operations. The. number of thermal and Icading-cycles used for . design ptirposes are 'shown in Tabic '4-8 of the FSAR. The maximum unit hcatup and cooldown rate of 100 F per hour satisfics stress limits for cyclic opera-tion. (2) The 200 psig pressure limit for the secondary side of the steam generator at.a temperaturc Icss than 100 F satisfice stress levels dor tem-peratures below the DTT.(3) The plate mate;jal and welds in the core region Lof the reactor. vessel have been tested to verify conformity t_o specified re-quirements and a maximum 1DTT value of 10 F has been determined based on Charpy V-notch tests. The maximum NDTT value obtained for the steam generator shcIl material' and welds was 40 F. Figures 3'.1.2-1 and 3.1.2-2 contain the limi ting reactor coolant system . pressure-temperature relationship for cperation at DTr(4) and below to assure -that. stress icycls are low enough to preclude brittic fracture. These stress ~ 1cycls and their bases arc defined in Section 4.3.3 of the PSAR. As a result of fast neutron irradiation in the region of the core, there will be an increase in the NDTT with accumulated nucicar ooeration. The predictcd maximu':t NDTT ' increase for the 40-year exposure is shobn on Figurc 4-10. (4) The actual shift in NDTT will be determined pcriodically during plant ooeration bytestingofigdiatedvesselmaterialsampleslocatedinthisorasimilar j c reactor vessel. The results of the irradiated sample testing will be evaluated and compared to the design curve (Figure 4-11 of FSAR) being used to predict the increase in transition temperature. The design y6ue for inst r.cutron (E > 1 21cv) exposure of the reactor vessel is 3.0 x 10" n/cm2sec at 256S MNt rated powcr and an integrated exposure of M n/cm2 co,. 40 years operr. tion. (6T The calculated maximum values 3.0 x IO are 2.*2 x 10M n/cm2sce and 2.2 x 10I9 2 n/cm integrated exposure for 40 yearr. nperation at S_0 percent load.(4). Figure 3.1.2-1 is based on the design value which is considcrably higher tha'i the calculated value. The DTT value for Figure 3.1.2-1 is based on the projected hDTT at the end of the first two years of operation. During these two years, the energy output has been con-servatively estimated to be 1.7 x 106 thermal megawatt days which is equiva-Icnt to 655 days at 2563.Wt core power. The projected fast neutron exposurc of the teactor vessel for the two years is 1.7 x 1018 n/cm which is based on 2 the.1.7 x'10' thermal megawatt days and the design value for fast neutron exposure. The actual shift -i'n ND'IT will be established periodically during plant opera-tion by stesti;a ve= sci material samples which are. irradiated cumulatively by securing them near the inside wall of this or a similar vessel in the core l area. To compensate for the increases in the NDTT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed -to stay within the established -stress. limits during heatup and cooldown. The NDTT shif t' and the nagnitude_ of the thermal and pressure stresses are sen-sitive to in'.cgrated reactor power and not to instantaneous power IcVel. Figures 3.1.2-1 a;;d 3.1.2-2 are applicable to reactor core thermal ratings up to 2$0$ M%t'. Amendment No.. ~ The presr.ure limit line on Figure 3.1.2-1 has been scIccted such that the . rcactor vessel stress resulting ~from internal pressure will not exceed 15 percent yicid' strength considering the following: A. A 25 psi error in measured pressure. B. System pressure is measured in either loop. C. 1 Maximum differential pressure between the_ point of system pressure .measur ment and reactor vessel inlet for all operating pump combinations. For. adequate conservatism, in lieu of' portions of the operational requirements ~ ~ f Appendix G to 10 CFR 50, a maximum pressure of 550 psig and a maximum o heatup rate of 50 F/hr (averaged over one hour) has been imposed below 275 F as shewn on Figure 3.1.2-1. The spray temperature difference restriction based on a stress analysis of the spray line no::1c is imposed to maintain the thermal stresses at the pressuri-zcr spray line nos:1c below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell. The heatup ana cooldown rates stated in this specification are intended as the maximum _ changes in temperature in one direction in a one hour period. The actual teaperature linear ramp rate may exceed the stated limits for a time period provided that the maximum total temperature difference does not exceed the limit and that a temperature hold is observed to prevent the total. temperature difference from exceeding the limit for the one hour period. REFERENCES (1) FSAR, Section 4.1.2.4 (2)' AS'!E Boiler and Pressure Code,Section III, N-415 (3) FSAR, Section 4.3.10.5 (4) FSAR, Section 4.3.3 (5) FSAR, Section 4.4.5 (6) FSAR, Sections 4.1.2.8 and 4.3.3 Amendeent No. 2, 2 0.. 'I ~, f.' ~IS-261' Item Component Exception - 6. 4 Bolting 26 Not Applicable 6.6 Integrally Welded Not_ Applicable . Valve Supports 412.3 ,The structural integrity of the reactor coolant system boundary shall be maintained at theslevel required by the original accep-tance; standards-throughout the life of the station. Any evidence, as a result of the tests outlined in Table IS-261 of Section XI of the code, that-defects have developed or grown shall te Jinvestigated.

4. 2.~ 4 To assure the structural integrity of the reactor internals throagh-out'the life of the. unit, the two sets of main internals bolts (connectingLthe core barrel to the core support, shield and to the

. lower grid. cylinder) shall. remain in place and under tension. This will be verified by visual inspection to determine that the welded r bolt locking caps remain in place. All locking caps willibe inspect-ed after hot functional testing and whenever the internals are removed from the vessel during a. refueling or maintenance shutdown. 1 The core barrel'to core support shield caps will be inspected each refueling shutdown. 4.2.5 Sufficient records of each inspection shall be kept to allow-comparison and evaluation of future inspections. '4.2.6 Complete surface and volumetric examination of-the reactor coolant pump flywheels will be conducted coincident with refueling or g-maintenance shutdowns such that within a 10 year period af ter start-i up.all four reactor ' coolant. pump flywheels will be exanined. 4.2.7 The reactor. vessel material irradiation surveillance specimens l removed from the reactor vessel in 1976 shall be installed, irradiated in and. withdrawn from the Davis-Besse Unit No. 1 [ reactor vessel in accordance with the schedule shown in Table 4.2-1. Following withdrawal of each' capsule listed in Table 4.2-1, Arkansas i . Power & Light Company chall be responsible for testing the specimens 4 and submitting!a report of test results in accordence with 10 CFR'50, [ Appendix H.. 4 i 4 < ~ i i 4 E Amendment No. 12,l29,, 77 p-i [. s + 4.2.8 The licensee shall submit a report or application for license amendment to the NRC within 90 days after the occurrence of any of the following: 1. Failure of Davis-Besse Unit-Mo. 1 to achieve commercial operation at 100% power by January 1,1978, or .2. Beginning one year after attainment of commercial operation at 100% power, any time that Davis-Besse Unit No. 1 fails to maintain a cumulative reactor utilization factor of greater than 65%. The repo

t. eball provide justification for continued operation of AND-1 v'

.e reactor vessel surveillance program conducted at Davis-7. N1. 1 or the application for license amendment shall pr.f ternative program for conduct of the ANO-1 4 reactor ves..;

2. ve111ance program.

Amendment No. 77a l l e j i l Table 4.2-1 ANO-1 CAPSULE ASSDiBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 CAPSULE INSERTION / WITHDRAWAL ~ ANI-E Has been withdrawn for testing Withdraw following 1st cycle at ANI-B Davis-Besse 1 ANI-A-Withdraw following 3rd cycle at . Davis-Desse 1 ANI-C Withdraw following 7th Cycle at Davis-Besse 1 ANI-D Insert in location WZ (upper) prior to 4th cycle at Davis-Besse 1; withdraw following 12th cycle ANI-F Insert in location YZ (upper) prior to 4th cycle at Davis-Besse 1; withdraw following llth cycle Bases The surveillance program has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code Inservico Inspection of Naclear Reactor Coolant Systems, 1971, including 3172 Summer Addenda edition. The number of reactor vessel specimens and the frequencies for removing and testing these specimens are provided to assure compliance with the requirements of Appendix H to 10,CFR Part 50. For the purpose of Technical Specification 4.2.8, the definition of Regulatory Guide 1.16, Revision 4 (August 1975) a) plies for the term " commercial operation". Cumulative reactor utilization factor is defined as: [(Cumulative thermal megawatt hours since attainment of commerical operation at 100% power) x 100] 4~[(licensed thermal power) x (cumulative hours since attainment of commercial operation at 100% power)]. ) i I Amendment No. 77b e I __