ML20065G440
| ML20065G440 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 02/28/1994 |
| From: | Madeyski A, Malone M, Zawalick S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20065G421 | List: |
| References | |
| WCAP-13931, NUDOCS 9404120274 | |
| Download: ML20065G440 (200) | |
Text
{{#Wiki_filter:_ - _ _ _ _ _ i Westinghouse Non-Proprietary Class 3- < I F Ij C y i i i e s 1 1 0 Ag , FDR f 4DOLF .
'. D e,' a c cg
.- - s .. . . i
-- -- _ _ . . _ _ _ _ _ _ _ _ _ ~ ^ ' ' ' ...: ..
.,s.... . , . , . - . . . . ..f. .
e.....c.. - 9
. -:_, 7. , > e......~.,y.,.- ,
. . . . . . m,
?b h., ^ ' ' '; ; i ' l, . .. a,. . [.'.... ' . . '.,;, ,
. , ;.. . .).eW$ Nf !'.:b f.i ';$. ,; g: .. , . . ._ y ,. )...},:'f:.&.N{. f :{, . ,. ..b l,: . ,;f, } :'.,, .. ,k. .[.*. .,..:
l . I_L .*; [' ' ' n llQ"..W.i ,- ll.. -.. :.. . ;-'. .; .. ;. -. ... . - . ...s ... ,.
. ... -<_.. . . <. ..' ..-....s . . . -1 . * . - - -
( ~: .-,.4g+ . m. M .. . .;.. .. A.A. . . t.> . :. .: . . ; .. ; r t, :h..,.a. .:; g...:. ,. ; . .L. ' :, ,:. ,.; _ . V. . . . ... .?... -.O..s. 3.- , , -
. . . . t . .. . .
.c ,:.
.-_ > .. , ,. .? . .
- r. y; . ;; . . : . s - :...; . ':..:, - ...'...:.:s .. ::. <-. .'.
%n. y l
- ....1
.; . .... a. - .:
s
;~
, L : ,j u. . ~ ..
q.ta..,e.;._ n:.y .f:.- ;
, .:L -.u
. . . ' ~. , .
- y , . 's $ ( .j.s.t.y:4' n
.. .ycJ . 7 3 . .u .y-:;;.,.j . ;': . Wf.y; .
.t .en .L .?.:
., ;J .h,:3:f
. , ~ , ., ; .) . if
- .:q .3,*[..,.w4 : 5
,;. .f.. s7'. p. .-. :,-:: ;..i.@.
< .. ,. . , 2., .
; . _ . , . .. i , .' ,; .-q i.;-,~;
..;-- -*- f.?-
;j '- -
- ,' . .m',. _ * -. .,: ;.....
- <'.,.., , . - . ,.i....t..(',. c'. . .:.
r .
-j . ' .. *. :. .. . ' .-> - '.
- qQ.*W
.' s e3 . .! > : .. .; .w- , . . ,,7*
s,... . ...x, . 's. -q , k;.-,. . I. . . . .
'4.,..'.-.. . ;. - :...,,.:
^; '.? . :.1; c'. ~5;/. ; . .:. . .,.,. .,y,z
. ; . e*L 3 e :... 41,- ,, ; +:.44,.
- g. v v;tg..':,.:.::' '..' .;,f, .,....f.,..lg.:, r'. .; ...f' %.,. ;.-, , F ,.,. ;' ;y7.'..s... . '
* - -,.,, . ' .y;.3. .:. :: . * - - -:-
i .= a;
- , . t , .
,, W,4.q:. j
.[.:::'-/ . *- - :~+c
. *. ('. ' ! '.,
- s. s. . . . -
;; r ;.:...s..x..
1 . . .
.c ,... * . ' ' . .
i:1 .h %. ..f.:(; 4 w: . . f:.? , $,;., '..x,.. J .i }';,' *;s .]' (q' . :.
,#:(
C . V;. .l .g " [;'f ;',..; . Y, ,J [
,[l} L; ;: . - '
- q. . ( . l - [ ' . :9.C -'. *' . .'l ,.," *
, f j. ?,
- .'c;; n ;;/ :. . . .:. .(m' - ~ . u, . -:: ': . ' :. ",'v....,. . . . < . .; ;.. ) . . ~ .a, . . -, :
s M. :c
, %, s -ir'
%. . s, ". % ,. m -s..p . .. .. .y
. ,c g
,b . .,.
,.8 .. . : :, ,.
, . u,p .; ;
s <
;.3.;...
. e- .
;',r- ' .t-h, -sj e g -s..
b.%.
' *...e *.:[,'M. ,"*.t.i S
. > , *.4
- ', t - h.e - @,;(Q,k 3
set .. g s <a J h .s.*. . ' . . - - ;.i ,[/'.f.~.;l,[';.v
- .cf
*e.a ,', p ..4
,; ; 3~,
e
; . .c , , .['
y>
, . _!h,N,9. . , , .. ,, , ,- :d, .[On,.
,, g ...: s,.- k ,.s.b(.(' 5.,:? ?%.;;$.
..,.3
,""J,. r
,.T,,
F. OJ. . 1[l[ t. . . .. i . ', .,
,: , h .-.' N , .- . * .[ y ;, .,, A ', .[;. ['th ! ,- [' E g ;j / , *. . ,I/ [,,
.), :' ', j*,f .} ' ,j, ', , , f , -
, x . c,) .$ p; .
.w,..- s l. r: :! <'-p .t . .a n ..f,.. . .: 4-%,b ::
.. - ., .r..
- 'h.: >p..
1-a3 *
,v'.j,.v,.,..
e .. m ..;. ,W Q .: - '..
,m. ,- .;*,.,, 3 : .,...n .\ e %t ... l.'.n .q . ,, . y e; y ' 4..,.s. Y..
.w . ,
- 6. - .
?.. .. +
1;yypf.IQ .:;. .'::. .W;] .R. ., : ' ,; -l.n.J..:.
. s 3, ,...,r,..
.%': ' :"Q,
.i . n - .:: . ;':; -";; _ +
. . u,k . .... .....; . ., . s.: t .
. .,. ~-
. $h . ,. - j. :. . ..kk'$ [I,[.1 . " I.f , . ~
['. V }.A.g;'l ..M,f. p .O '..b s.:- -t; .: . . . . . . : ; ., e :n .n;,._ , 1 ,.
,g.' ..- . ' -- . . ' ' . . ' . . '..
_.m. . .:.
; ; %:'..: ~2.. r..; *: : < . w. 1
.;.;r.;h: . ':... ::, :;3
,, 4, 7,7 :
.y 4A :
<:L ...,- l :' t .a. ....~.m.*:.,.. , ;; :.:.E..d.. Q.4; . . , . . . . . .. . , , . ..;.
.u-t., s . :: . . i+ ..:.,e' . f: .v. .. f. . n;>.s..e .. . n.. K... e.. - -
- y. e ,'.:,. .* .;-:.: ' ,. ,.a s .
.; .. . . :. ..M. * . .-
;,,: ' , ' ..a"_
, , _ ;v .: - ; .1 ' ,
. f.f ' .s, _ l.. . ,?
. ; f . f - * :y . .t,.-, s. . . _ .' . ) 5:
.> e.. > ; 3 y
m4 .2i-. ~ . . -
- r. +....,..g,,. ,
a g. ,
. > .a .-
D Me c.: l# 9. ."i;$b, M~...c=fik.
.? ?.:I h
- ,- - - . R, 4.
....s
..V h4
. y' . d. i [; ..,R) p-. . .;. < .. D. ;s, )72 - "
[ . . '.
,J f. ,
4; :, h,,*,/ ., . , a ,.:. .,) .i '.f7.'.[h. . . ]c. .I , [3.e ;j T.v.d..[t g.M ,2^ l . ._ ( f '; , ,? .
..y v. -e 3. g
#e- .,JL..h N.: :,,:,.~..eF .; o
.c....
...e.v. : < , .rs. .. . . f g : . . . . ....:f m y;%... ... . ... ,. . . .
- 1. %~ y ) Lp.
1
; pem'. m c. .o. ;,e J. \ *:,;; W ^ :': .. x..r. .
?-lL s . . ~ -
- .* '; . .. ..w L0 . a...5e. ,*., . *: m. : . . v ? :.;tnV-
;.g:?x (l. ;m. _
., y :,. - ... y - ;_ . y . .. y*.y.5s: f.A'y.ht '; :... .e. .. .
p 5. z,w -
.+;- ~ .. $., .
Q p i.,
.:. Q., .. ' 9 's. f.; F.T,3 g ..:' '.' , . . - , 4;g A. :s
. , ';..: ;. y ,,.?,. f 3w l,4 d . .
& .~..& c W ' :
4 :,
.: 7), ; . : .-
- v.1 C. .H: . t : ..b . ,7.
y' .t . i =V?:. y&,.;. l^..0r:J.': n
. .'.g. ! ? , y.y;.-h'h. %'h '/;? , - ^, f. . e. , ? .. . -
;; ,:2 . v '.;.2:*; :- ; g'$'.r:ing;L '
- y. , , ..
.. :} {*D.N. : . ; .:, .. : . h,s "{, :
.?..'.. !. . . <: .?. ,.:.j%.: - , o .y; f . ; ' K. c y.h;+f 3;> . , . >.. w - .:~ ,%. ,y: ..q:v.,
. .; :.,; e ; .. . .,. - b. 7 i
- 9- C . :. 7 % :. .w ; , , 'a -:, *. $ f ' _ ':.. .r . .i .. J . ' , _ l .Y m.Y' .,> .:, ;.L .' ' ...,c ,-. '. ....t..
- 3.. *.g?w ,s e. %. .. s , c. , - us. :. #..?*... ...s ..c.v. ;
.t.c . : : w. ::*-
- :s, - . . .. ,. .. .- .-
. .. n's.3. 4sw.. ....3.. .: ': .: .
.'Ih....
x;p
.fr . m.5g' ? ' l- ,. .
. 'T0: ,_, *) .l f *f_ j i ; a : '. ?l;l .: 'Q ' '..
-.r '
-l'.,~_. .%'- /.
.x ^ -< ~ ' ' '.
{Uy, .;. .s:*:{'. .f G,l:: .ly. .:(. l [ :.. ,.! . .. .; *- ,5 -{;3, ,: ': a*..tK ._. :;#.:- ;_. ,:,.' i'...,:. - ':; ' " . '::: .. _ ' - ;
.',- ';; : \.. -. ',.'. ._'l,_.
?.t... U,( . .;}:.;'. ;; _ :;.+ .',,s;.p - ;'..y-- ;' ,'
.: e f.
4 . ; .s _, ,
,'.'.,.?.s..,. s ._s .
..'..-:' ..:- ; r .. '.
. .. .,e . . .~-. - .p
. y . , :, : * :, ..
4 ;,. r' ? ~. '2 ,y y, '. ** pw g;.- f.y;: ' :.
. n;.v.y t n:.,s v:: .F ?
, .,'.i.*sr-._,.'o,;'
. .n ,
',./- l 5?'., { .
'.,'.\'
. . , - ' . . - . , * .c
#, *b; ; ,,' . ; ,,...'...{9 fvv :;.c .<j. : .'. ..:. ' . .;!,': ;n., q q, ',.
h .*: . .b ..h ,
- 5 ,, ' ' b- - - * '* . . .,' . * '
- f. . t .-' .- [*.
,'.;._..!m.
; ':;- *e :
' . : ?-
.. . . ~.. . ' . . * *.. .. ..* .. ' . '- .. "-- ..i.
s
" , . ..._ ;. .': .i- r : . - *-...:.. ;;__y' - . ' .-- .** . . .. -
l .; ;3..; y . f,,u2.
., ,,<a...:.,,,.,....,...;:<....c.,.,,.j-
,:3 . :: n . . :.. - ,t ,. .
"p
< $. t : . T.Q: j.g. j,g' ; , (, . . ". '.f, -
g W,-('r :.. , .;~."det". W... .)se.y.' ' ), i~':) d. q ..W.. g.,l3, r. */IV .. ,ec . 'y L U .' .. '. . ' .?;'. i . ' .Y......'.'. ' * '
., ..,_ .. ... , . , .. .i ~. . .. '.. :. ' . ' " , ' . ,l '. '
',,.s i '. .. . ;.. f.,.y . '., f /.: .g. g : s..d. '.. *: :.$,3. :.;.:. - s .* M< ,Q . . . . ;. A. ;. i . . , , , .l,, , :. , ',i' jf ' . ' ' . :.
7 ,(s.g. 4 . . .. .* , . . . ' , ..' ' . ' . . . ' ' - ' .. . ' 7 i ) .i . g % . ,, j.q,Q.. .# ,u . 4, j 7 5. :.f - , ' , 73 +,
.- .,; c r - .;. . . . g.,'.. -:-, ....,j s
.- . p
. . ..' _. ,.:, ,g. g * : :.-
, . ;- - . . , -. . . . . q
..v v' : ; .g. . ;;;fi;W.. ;;,...; n \ ,.~ .,. . -
-y ..-..i. ,..,g,..n..:<.:,;,;;_ s g . . ,.; s . e..r.ff
. _ ('* ; .
+... ,,...C.;,.... .s.f.. . ; ,. : ,..,..e.. .- *..
,. . ; . ...- : c <.. ,. . ..
.r 3
.?,. T - .
5;... '. - ' ' c. '.......:.
' . ' ~ . .
X.. 3.. . s. .".. '.'
. ?
..; .: ,.,. . :. :. . ..~. , - . . . ', .. __ _.. ., .. -_ .,_~,, . . . . . . _ .,.
q: w, :3. .:i( ...,,.,'.'.%.s 1
. . . . ? . i. .;
c. s f. .zo.r;; . .,e %<,,t .,;; *. :;.> p ..;" s .t,; o..y ng: L.y..:
.. .> .c : ; .n ' -
y _ . - ... ;..c ,- .. . . ' . f.4a .:. . .- .... . . . ,., . r ..;,. .. ...._%
- c. .
. . 'y f' :[ ' '-p,s ... gcp. . :.. . ., :: * , f, T. .; ', : l., $.. - ".-;: ' ' ' '
Ti . . . . e-Q: ' , r ,..p W:'..:*4 '\.' * . 5.
' ,\
i * . -.-= '.- ..- .: '*
.n) ,.
i
. ;. M . ' . . ' . . l.* ;' :_ ::' . . ' -
. s
., ;p: $ r'.?: r. .; . ;..y' :. .;; Q pc ;n ..g . ; : ., ,; _q ~; ::;,7; ;'; .jl[y. g.:
. i:.';: . . . .*:.- . * ; _. :s ;
4"
;- w ':.. .......'.j_..-.: -.
, - . , ; . u 1:;:,. n., y.'.,. :. .::
? .e.
G.v.v._
~n .y _ .,r r..A:.; -.,c. ,...l,.
.,.,2.. : A :..ht u:#.-- . .-.>r.. .'.Q.,,:..l*".- . .- ." ._ . ,_ .-,;- . - - ; '. . ?
c" . . . - . Qf.y,e :u t.n 2 i.g. ';jk' ':;,lz. &*g . y f.: ;s
. ? - ~:.:
7
- n'.*-
;. :-: . ?
.&~
- 2 g.,
x . .f). ' . T :.y..2;u . ~.. f;:: 9.%,..
- ~
9 s
.. ;:& :;.yJ: :4
.: n.'
. y sh.
s
- :._ 9 ;;;..
- * :L. ,
~.g - .. . . . . . - .: *-* .'..: . * - vs. ; ..;,.
. - -" .'.' .. 1.. .'.'.;
.. .> *'.v.. .
", .:~. ]
; O .;'. l g[:.~;';): :. J;.g' y,
.: 1
.[I % o n :.n;;,;' .3. 5k*j, (r.",h..f~T :l,.dy
,o.
.n ,p h.?
r ' ,'*. :'l.1Jc.~ ,"; 4
- s. _. ; [;,' ,1 lI
- J ,# I '<'::,;.. 1;j e .-
.. ..: %'.[ :. .. .[.
~.d 2.-
,,, ; \ h $ y .: . ,{:
. . . . ~ j,.
h' f mf h, .' N,' rY ' f.. l 4 l .;:l 1::..": . . . . . . . .l'l D .' .' . . l, f, . * . \ .' ' *.>
'".'Y ,';.... ,;' .
-? . '. >;. . ' . Y. . q.
"., ll* :, >. : ? ? - * :'..a'.. ^
..T.
/..f : ' ' ' ':? Y; ,Q d.f. LO:! qis .;k,*
.; n
?:
- .A t%4 & . d,' : G^':[:N,Q:%
L l.l:%ql*
- . L ' , . R ' . .; ' M%. <;
~:.' - * ' -" .,,;
${f. . M J h ?
.'.Y'.*'
': ','.- :: ,./..'- ::_-. ,: ' , $ .'. .Ml,f.,:n_:l:,., ?
V, :h:g ";;' e
< W c;: , R'" * ;.;;';;:.t ~. .' '! -. ~ :;.;.l .d .. c '_: . . ' .r d '. y ' , t, o.. .& ..'*; . . -- .' : '
.'i . .
.nWW ::.. ' .> :- ~ * *.
- f. .h 'l: - l...%..,,'. ' 's
. , ?. i:.; h.yV,f..% , r !' . :- - %..*a,.
N:n.'.lg_!. .: L.' . ',j,
- l '.... . .: wse -J' .: - ~. .-:-' '.. . ' \ ;: * .- - -
. Or.er.,e
- .jn. .e,
's n . . ?, r,-.n.do7.., ..t. f.. . .c..H. +e.a.:::..;, ..; . .?. ".+:.w. .- 1;. . .. . . . . . .. ....s : . .
yn - ~- s , n. va g, c. ? w . k, .. y,j-pf.y' ;_ g% *?...;r.Y... ..n:: .fi *
~:;.-a....,...:,". : . .. -. ,..'i.' :. - . ...::..-. -.t r
- 4. ' ....;. g . . . .. . , . .
. . . . . ~ . , _ , . , . .
1.n> ~. a.,, * ':Ql. ,s v.:9.;m ? :,' . -% i f .'..* - ~ ,'~.' I' + ' r.) ~ _
. s 3-
.r A<!.f: c .f ..'!f : f. C:* l ,'*Q . ,.>.' ~: :\ ' ' ' ' '
- M
'; ' ~ - ' , .' * ..* - ' 'd GiffM.\.",f.$ h l1,L'(:.d:".'g *.'..J.".*"~' c' d ' . - .' '*O l
i, _ -l' ;.:sj$%':' $!4:*n
' ^ ~ ' ' ^
^ ' '
J- "\
- .:;.b ':;. '.c' :"....-f ::" i ,,; '. ",' . ' s' % ' '. ;,.] .:;-. '"- ' ' ..',:, . ' . ' :- i: ' ., ..
;l p_ T.fi'5.% f. .%:
1 :..M.M '.' .,~ ') J,.l.f.i~.klh,*,.k.M j . b., : 5 'P.; . ~ :'. ll*.',:c;l.x,:; ,
.?- .
's '.
- ...:.'.'*.'s..
.f. 9;, ,y , . 4.,5.;.). . :;. . a.:
h - : . '
., . . , ~ . . . .. ? . . ;
o..?. k.i pq: . .
. . :. ' ..: * './,f f.g 's,- A, ' -l W, .:. : w.. .s :.e .'. , . .
3c: r. .s. - : *r - ;. fy - t...s,. .. ..;. . - s' ::. . .t
; &,, g .. l ~~'%.
p s . . . .v, h.:t.i , * .e .C. ._ .y,W., "%' i:, ',' * \. . l ~,. L. .c '!. .t.W.,
.. p . t g
.l l.V .* '. :( ; ' . J:. .' s.. . :
~
.F-.
L :." . - 5.. . & ~ - .
&( t,l.'.[ *_ . ' .*:,Q . :,h: ::';. s. .o
.s ' .,S.,d;Wy.!.U
. , :. , ~
/r 2 ,
. *Q . ,;i; ~ j'[. '} 'f;.f. of . ': : M ' 9 ' .;. ? .: .
- s . J
?_ ' ;, . ~ .
- T. 7..l *.
- 4 g m -o. v. ' o- n % ' : r: < w . i. . . : .. e - . -
% .s ; . : < . ' ' -' - . .
..4 :fW i.1 }9 's\.': 3::.a -W).Ni;*s&,2. l.
. ::*0,:.Q W' N.':,V .:.U ?l. :"'),Q' O 'L .e.:- L: .\. < :..- '
- . .'. :' ** .' T. 4 . .*
?inf: .*7:. . ' < ' . . - ?; ..Y_
c.
.l. M L: ' : .
h .i :" - y
.. r: ? ,' .- . . ., . .
~. ..- a
. $ .. - * . . ..:. ~
- ~ :
' .;'. -fy. $.<. ..e_
v :s. : ..s; a . -::[ *,J?.!:;p.g?: y:.. v.Q.'.e. .L: .',,e'y:)' . .h;;;' . ;, ; 4 . . ".".;t ' < .* *
- 0. . sht .' ;; ' . ':: - '- r n, ',
, >'-. + L.? - _,Ql *.
soi. };?s , 4. c.s:,. <.. .z,' vs f' _;.
. ;,^..g.. . ?a,. .;.: , :y .e.h,:
w -=: s. .c o
- . . . p. .*;;y .: ^ *: .r.
n ..
. . - ~ .. .y' ' ; .
'. . ;. - ' ..a,M... ,. ' , . .
- .,: -t : -
m [s f ? ; ,h; ; +. .:,al. e '- . ' . . ' ' . -
. :. ,:.V.L:.% L,.:c;s ::'-2.' ':. .1 % : . . . ..,x .:. '8.s. . . ._ . ':. . .
, . - - . :. ':2 y ' . ; __.....: :
L Q)>c - g . ; . . :. .: c.;p. 'y;w. n
+ n .,. ,,w. - .,; t.5 ; .. ..
i
.s .. . . . . \..: .;.; ,; n . :
.c. ....;
. ~ *
, + V..s n.s... >,.;.:..,. ;...
.......i..- ...,';.
,. y ,e . .g . ; n. 4.uos4.: . _ , .; q. .; . :.s>.. g , . ; ., ..:,_ .y...re,. . ' ,: . . . >- m.. .q.,-
- . - _. . , . u.;-
.->-s. _ a : -.
- - *: . sg,..7.e .. *.\p. . - . - .-
s..
% .p-O -
*.$. w. A M..~ ..;..' > ; e n. 7,./...*c....u.d" .w , ,
;- .' n.
w . : . . .. - - ;- , ,"
'- '. . a '- A ..,' '
'p '.
[* 4? g,[ M'.a ; ;, . '.f:L .'*E'
* . f :' .
- ~ ,. . : . . ~- .: . .. ' . . - .,' '.. a. . : : , '.:W L : '
. &')y' f. %;,;):T: . {. '. :) g. , . . .. . : f. .' :' L, .L8 F. l:n + Jn '. j ... '. ':.< .....-..
V ;
%, ' x -:l': y: '..; 4. <l.Q'.h. . . v.. m ..vl~ .;. .:. .. .. . .
,.* . [ 5 "[ .* Lp * , , - . .** I .'A } *: ,g -
' .g.f.g . .. ...;':<'a..v.'..,_,,...'.
g
'y. , ;: - ; ;- ,-:' : .. : ~..: . , ; . v. .. * * ;:, . . . ';
.." 'u. .. . . */ -~.,9 e e, .:w.- .: . u. : . . .;; ... ,,, .9 w; ,qw. :-;,,p m;. $p,cy,*M, t
k,, ,g.-
- . ; n
.. e ~ ;. " . . ;
., e - .-. .
i l. WESTINGHOUSE CLASS 3 (Non-Proprietary) I i ! WCAP-13931 i
- l. -l' r .
a j h 1 ANALYSIS OF CAPSULE Y FROM THE , GEORGIA POWER COMPANY VOGTLE UNIT 1 ' REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM M. J. Malone S. S. Zawalick
~
A. Madeyski '
- i. .
l= l February 1994 i Work Performed 'Under Shop Order GTXP-106
, Prepared by Westinghouse Electric Corporation for the Georgia Power Cornpany t
! P Approved by: bb_, .wJ ~ l T.'A. Mey er, Managed Structural Reliability and j Plant Life Optimization J WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division l
, P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
@ 1994 Westinghouse Electric Corporation l
l PREFACE l This report has been technically reviewed and verified. Reviewer:
" c Sections 1 through 5,7,8, and Appendix A E. Terek Appendix B B. A. Bishop L ,o) ;
'/ I \ .
Section 6 E. P. Lippincott
/
O 9 i b l d I 1 0 e i
i i i TABLE OF CONTENTS Section Title Page . l 1.0
SUMMARY
OF RESULTS 1-1 i
2.0 INTRODUCTION
2-1
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE Y 5-1 5.1 Overview 5-1 l 5.2 Charpy V-Notch Impact Test Results 5-3 l 5.3 Tension Test Results 5-5 5.4 Compact Tension Specimen Tests 5-6 i 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 l . 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-6 6.4 Projections of Pressure Vessel Exposure 6-10 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 l l 1
8.0 REFERENCES
8-1 l APPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS l APPENDIX B - HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL , OPERATION OF THE VOGTLE UNIT 1 REACTOR PRESSURE VESSEL - l ii I
1 I i LIST OF TABLES l' i I i- Table Title Page i 1-i'~ 4-1 Chemical Composition of the Vogtle Unit 1 Reactor Vessel- 4-3 Intermediate Shell Plate B8805-3 i 1 4-2 Chemical Composition (wt%) of the Vogtle Unit 1 Reactor Vessel 4-4 Surveillance Weld Metal ] 4-3 Heat Treatment of the Vogtle Unit 1 Reactor Vessel Surveillance 4-5 i Materials i i ! 4-4 Chemi.try Results from the IAw Alloy Steel NIST Certified Reference- 4-6 j Standards l l
- 5-1 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate 5-7 l- B8805-3 Irradiated at 550 F to a Fluence of 1.24 x 16' n/cm2 l (E > 1.0 MeV) (Longitudinal Orientation) l 5-2 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate 5-8 l
2 l B8805-3 Irradiated at 550 F to a Fluence of 1.24 x 10" n/cm (E > 1.0 MeV)(Transverse Orientation) 1 - 5-3 Charpy V-notch Data for the Vogtle Unit 1 Surveillance Weld Metal 5-9 2 t Irradiated at 550 F to a Fluence of 1.24 X Id' n/cm (E > 1.0 MeV) i 5-4 Charpy V-notch Data for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) 5-10 2 l Metal Irradiated at 550 F to a Fluence of 1.24 X 10' n/cm (E > 1.0 MeV) l l, 5-5 Instmmented Charpy Impact Test Results for the Vogtle Unit 1 5 11 Intermediate Shell Plate B8805-3 Irradiated at 550*F to a Fluence of l- 1.24 x 10 n/cm2(E > 1.0 MeV) (Longitudinal Orientation) ! l l '.; J i iii 3 l
LIST OF TABLES (continued) Table Title Page - l
- i 5-6 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 5-12 i Intermediate Shell Plate B8805-3 Irradiated at 550 F to a Fluence of 2 2 l 1.24 x 10 ' n/cm (E > 1.0 MeV) (Transverse Orientation) 5-7 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 5-13 Surveillance Weld Metal Irradiated at 550 F to a Fluence of 2
l.24 x 10 ' n/cm (E > 1.0 MeV) l . 5-8 Instrumented Charpy Impact Test Results fcr the Vogtle Unit 1 5-14 Surveillance Heat-Affected-Zone (HAZ) Meta' Irradiated at 550 F to a 2 Fluence of 1.24 x 10 n/cm (E > 1.0 MeV) 2 5-9 Effect of 550 F Irradiation to 1.24 x 10" n/cm (E > 1.0 MeV) on the Notch 5-15 Toughness Properties of the Vogtle Unit 1 Reactor Vessel Surveillance Materials
~
5-10 Comparison of the Vogtle Unit 1 Surveillance Material 30 ft-lb Transition 5-16 Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-11 Tensile Properties of the Vogtle Unit 1 Reactor Vessel Surveillance 5-17 2 l Materials Irradiated at 550 F to 1.24 x 10* n/cm (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center 6-14 I 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates at the 6-15 Pressure Vessel Clad / Base Metal Interface , [ 6-3 Relative Radial Distribution of $(E > 1.0 MeV) Chin the Pressure 6-16 - Vessel Wall iv
i , l LIST OF TABLES (continued) l !- Table Title Page l i l l l~ 6-4 Relative Radial Distribution of $(E > 0.1 MeV) within the Pressure 6-17 Vessel Wall 6-5 Relative Radial Distribution of dpa/sec within the Pressure Vessel Wall 6-18 6-6 Nuclear Parameters used in the Evaluation of Neutron Sensors 6-19 67 Monthly Thermal Generation During the First Four Fuel Cycles of the 6-20 l Vogtle Unit 1 Reactor j 6-8 Measured Sensor Activities and Reaction Rates Surveillance Capsule Y 6-21 Saturated Activities and Derived Fast Neutron Flux 6-9 Summary of Neutron Dosimetry Results Surveillance Capsules Y and U 6-22 l 6-10 Comparison of Measured and Ferret Calculated Reaction Rates at the 6-23 Surveillance Capsule Center Surveillance Capsule Y 6-11 Adjusted Neutron Energy Spectrum at the Center of the Surveillance 6-24
~
Capsule Y 6-12 Comparison of Calculated and Measured Neutron Exposure Levels for 6-25 Vogtle Unit 1 Surveillance Capsules Y and U 6-13 Neutron Exposure Projections at Key locations on the Pressure Vessel 6-26 Clad / Base Metal Interface 6-14 Neutron Exposure Values at the 1/4 and 3/4 depths into the Reactor Vessel 6-27 Wd v
1 1 ! l j LIST OF TABLES (continued) ; 1 Table Title Page . l 6-15 Updated Lead Factors for Vogtle Unit 1 Stuveillance Capsules 6-28 '1 7-1 Vogtle Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule 7-1 t i l l l
- i 1
i l O Vi l l
l l LIST OF ILLUSTRATIONS ,- Figure Title Page l l l~ 4-1 Arrangement of Surveillance Capsules in the Vogtle Unit 1 Reactor 4-8 l Vessel l 4-2 Capsule Y Diagram Showing the Location of Specimens, Thermal Monitors 4-9 and Dosimeters 5-1 Charpy V-Notch Impact Properties for Vogtle Unit 1 Reactor Vessel 5-18 Intermediate Shell Plate B8805-3 (Iangitudinal Orientation) 5-2 Charpy V-Notch Impact Properties for Vogtle Unit 1 Reactor Vessel 5-19 Intermediate Shell Plate B8805-3 (Transverse Orientation) l l 53 Charpy V-Notch Impact Properties for Vogtle Unit 1 Reactor Vessel 5-20 Surveillance Weld Metal l
~
l 5-4 Charpy V-Notch Impact Properties for Vogtle Unit 1 Reactor Vessel 5 21 Weld Heat-Affected-Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor 5-22 Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) ; 5-6 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor 5-23 Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) 57 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor 5-24 Vessel Surveillance Weld Metal l-5-8 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor 5-25
. Vessel Heat-Affected-Zone Metal vil
l l l LIST OF ILLUSTRATIONS (continued) l l Figure Title Page - I i 5-9 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell 5-26 ~l Plate B8805-3 (Longitudinal Orientstion) 5-10 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell 5 l Plate B8805-3 (Transverse Orien:ation) 5-11 Tensile Properties for Vogtle Unit 1 Reactor Vessel Surveillance Weld 5-28 Metal i l ; 5 12 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel 5-29 Intermediate Shell Plate B8805-3 (Longitudinal Orientation) 5-13 Fractured Tensile Speciraens from Vogtle Unit 1 Reactor Vessel .5-30 Intermediate Shell Plate B8805-3 (Transverse Orientation)
- l 5-14 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel 5-31 Surveillance Weld Metal 5-15 Engineering Stress-Strain Curves for Intermediate Shell Plate B8805-3 5-32 Tensile Specimens AL13 and AL14 (IAngitudinal Orientation) l .
. 5-16 Engineering Stress-Strain Curve for Intermediate Shell Plate B8805-3 5-33 r i
\
l Tensile Specimen AL15 (Longitudinal Orientation) 1 5-17 Engineering Stress-Strain Curves for Intermediate Shell Plate B8805-3 5-34 : Tensile Specimens AT13 and ATl4 (Transverse Orientation) l 5-18 Engineering Stress-Strain Curve for Intermediate Shell Plate B8805-3 5-35 Tensile Specimen ATIS (Transverse Orientation) - viii i
1 l LIST OF ILLUSTRATIONS (continued) l l l Figure Title Page ) l-l 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens 5-36 I AW13 and AW14 5-20 Engineering Stress-Strain Curve for Weld Tensile Specimen AW15 5-37 ( 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-12 1 l 6-2 Axial Distribution of Neutron Fluence (E > 1.0 MeV) Along the 6-13 l l 45 Degree Azimuth l I l i-M L l !~ I
- 1 i
l ix i
-m- , ,v ,
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule Y, the second capsule to
~
be removed from the Vogtle Unit I reactor pressure vessel, led to the following conclusions: 2 o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 1.24 x 10 n/cm after 4.6% effective full power years (EFPY) of plant operation. o Irradiation of the reactor vessel intermediate shell plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation), to 1.24 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 40 F and a 50 ft-lb .ransition temperature increase of 35*F. This results in an irradiated 30 ft-lb transitic,*1 teraperature of 25 F and an irradiated 50 ft-lb transition temperature of 55 F for the longitudinally oriented specimens. o Irradiation of the reactor vessel intermediate shell plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate 2 (transverse orientation), to 1.24 x 10
- n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition
~
temperature increase of 20 F and a 50 ft-lb transition temperature increase of 30 F. This results in an irradiated 30 ft-lb transition temperature of 35 F and an irradiated 50 ft lb transition temperature of 95 F for transversely oriented specimens.
~
o Irradiation of the weld metal Charpy specimens to 1.24 x 10 n/cm2 (E > 1.0 MeV) resulted in no 30 and 50 ft-lb transition temperature increases. Thus, the irradiated 30 ft-lb transition temperature remains at -40'F and the irradiated 50 fi-lb transition temperature remains at -25'F. o Irradiation of the weld Heat Affected-Zone (HAZ) metal Charpy specimens to 1.24 x 10t ' n/cm2 (E > 1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 25 F. This results in an irradiated 30 ft-lb transition temperature of -50 F and an irradiated 50 ft-lb transition temperature of -25 F. 1 1-1
i l o The average upper shelf energy of the intermediate shell plate B8805-3 (longitudinal l 2 orientation) resulted in no average energy decrease afte.r irradiation to 1.24 x 10 ' n/cm (E > 1.0 MeV). Thus, the unirradiated average upper shelf energy ci 122 ft-lbs for the - longitudinally oriented specimens remains unchanged. t . I ) o The average upper shelf energy of the intermediate shell plate B8805-3 (transverse orientation) I 2 resulted in no average energy decrease after irradiation to 1.24 x 10 n/cm (E > 1.0 MeV). l Thus, the unirradiated average upper sitelf energy of % ft-lbs for the transversely oriented l specimens remains unchanged. o The average upper shelf enagy of the weld metal Charpy specimens resuited in an average 2 energy decrease of 1 ft-lb after inadiation to 1.24 x 10 n/cm (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 144 ft-lbs for the weld metal specimens. o The average upper shelf energy of the weld HAZ metal decreased 12 ft-lbs after irradiation to 1.24 x 10 ' n/cm2 (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 2 124 ft-lbs for the weld HAZ metal. o The surveillance Capsule Y test results indicate that the surveillance material 30 ft-lb transition temperature increases and the average upper shelf energy decreases are less than the Regulatory Guide 1.99, Revision 2 N predictions. o The calculated end-of-license (32 EFPY) maximum neutron fluences (E > 1.0 MeV) for the Vogtle Unit I reactor vessel are as follows: 2 Vessel inner radius * = 2.176 x 10 ' n/cm 2 Vessel 1/4 thickness = 1.188 x 10 n/cm 28 2 Vessel 3/4 thickness = 2.568 x 10 n/cm
- Clad / base metalinterface j
~
i l l I l-2
l l SECITON
2.0 INTRODUCTION
l' This report presents the results of the examination of Capsule Y, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron l irradiation on the Georgia Power Company Alvin W. Vogtle Unit I reactor pressure vessel materials actual operating conditions. The surveillance program for the Georgia Power Company Vogtle Unit I reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillana "mgram and the preirradiation mechanical properties of the reactor vessel materials is presented in W 4E 011. " Georgia Power Company Alvin W. Vogtle Unit No.1 Reactor Vessel Radiation Surveillance Program"m. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels"N. Capsule "Y" was removed from the reactor after less than 5 EFPY of exposure and shipped to the Westinghouse Science and Technology Center Hot Cell Facility, where the postirradiation mechanical testing of the .py V-notch impact and tensile surveillance specimens was performed. i This report summarizes the testing of and the postirradiation data obtained from surveillance capsule "Y" removed from the Georgia Power Company Vogtle Unit I reactor vessel and discusses the analysis of the data, i I i 2-1
SECTION 3.0 i BACKGROUND l l The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron inadiation on the mechanical properties oflow alloy, ferritic pressure vessel steels such as A533 Grade B Class 1 (base material of the Vogtle Unit I reactor pressure vessel) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation. A method for performing analyses to guard against fast f acture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel CodeH1 The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTypy). RTxor is defined as the greater of either the drop weight nil-ductility transition temperature (hT)TT per ASTM E-208m) or the temperature 60 F less than the 50 ft lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented nornial (transverse) to the major working direction of the plate. The RTyor of a given material is used to index that material to a reference stress intensity factor curve (Km curve) which appears in Appendix G to the ASME CodeW The Km curve is a lower bound of dynamic, crack anest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Km curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors. RTyor and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, such as the Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program m , in l which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. 3-1
/ The increase in the average Charpy V-notch temperature (ART xo rat 30 ft lbs) due to irradiation is added to the initial RTxor to adjust the RTw or ( \RT) for radiation embrittlement. This ART (RTxor initial + ARTxor) is used to index the material to the Kmcurve and, in turn, to set operating limits for - the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials. 1 I I I i I l l l I I l l 1
~
l I i i i 3-2 l
SECTION
4.0 DESCRIPTION
OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Vogtle Unit I reactor { [ pressure vessel core region materials were inserted in the reactor vessel prior to initial plant start-up. , The six capsules were positioned in the reactor vessel between the neutron pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. The capsules contain specimens made from intermediate shell plate B8805-3 (Heat No. C0623-1) and weld metal fabricated with 3/16-inch Mil B-4 weld filler wire, heat number 83653 and Linde 0091 flux, lot number 3536, which is identical to that used in the actual fabrication of the intermediate to a lower shell girth weld and all longitudinal weld seams of both the intermediate and lower shell plates of the pressure vessel. i Capsule Y was removed after 4.64 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2T compact specimens (1/2T-CT) made from intermediate shell plate B8805 3 and submerged are weld metal identical to the closing girth and intermediate and lower shell longitudinal seams and Charpy V-notch specimens from the weld Heat-Affected-Zone (HAZ) of intermediate shell plate B8805-1. i l^ Test material obtained from the intermediate shell plate (after thermal heat treatment and forming of a the plate) was taken at least one plate thickness from the quenched edges of the plate. All test specimens were machined from the 1/4 and 3/4 thickness locations of the plate after performing a simulated postweld, stress-relieving treatment on the test material. Specimens from weld metal and HAZ metal were machined from a stress-relieved weldment joining intermediate shell plate B8805-1 and adjacent lower shell plate B8606-3. All heat-affected-zone specimens were obtained from the l weld heat-affected-zone of intermediate shell plate B8805-1. Charpy V-notch impact and tension specimens were machined from intermediate shell plate B8805-3 q in both the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction of the r e) and transverse orientation (longitudinal axis of the specimen normal to the major rolling direction 02 the plate). Charpy V-notch and tensile specimens from the weld metal were oriented such tia, the long dimension of the specimen was normal to the welding direction. The notch 4 - of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction. 4-1 4
Capsule Y, also, contained 1/2T-CT test specimens from intermediate shell plate B8805-3 machined in both the longitudinal and transverse orientations. 1/2T-CT test specimens from the weld metal were machined such that the simulated crack in the specimen would propagate in the direction of welding. - All specimens were fatigue pre-cracked according to ASTM E399.
~ l The chemical composition and heat treatment of the surveillance material is presented in Tables 4-1 through 4-3. The chemical analysis reported in Table 4-1 was obtained from unirradiated material used in the surveillance programNand irradiated material from capsules UN and Y.
l l l l Capsule Y contained dosimater wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmirm shielded dosimeters of neptunium (Np*) and uranium (U*) were placed in the capsule to measure the integrated flux at specific neutron energy levels. l l l The capsule contained thermal monitors made from two low-melting-point eutectic alloys and sealed in 1 Pyrex tubes. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two eutectic alloys and their melting points are as follows: 2.5% Ag,97.5% Pb Melting Point: 579 F (304 C) 1.5% Ag,1.0% Sn,97.5% Pb Melting Point: 590 F(310 C) The arrangement of the various mechanical specimens, dosimeters and therTnal monitors contained in capsule Y is shown in Figure 4-2. I
. 1 4-2 l
m l l TABLE 4-1 Chemical Composition of the Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 l Chemical Composition (wt.%) Westinghouse CE Analysist2) Capsule U'd Capsule W Element Analysist23 Analysis Analysis C 0.220 0.250 -- 0.225 Mn 1.320 1.320 1.262 1.277 P 0.017 0.003 0.010 <0.015 S 0.011 0.010 -- 0.0139 Si 0.280 0.260 -- 0.232 Ni 0.610 0.600 0.586 0.584 Mo 0.570 0.530 0.431 0.527 Cr 0.057 0.040 0.049 0.057 Cu 0.058 0.060 0.053 0.061 A1 0.030 0.029 -- 0.032 Co 0.006 0.009 0.013 0.008 Pb <0.001 <0.001 -- -- W <0.010 <0.010 --
<0.037 Ti 0.004 <0.010 --
<0.008 Zr <0.002 <0.001 --
<0.009 l I
V <0.002 0.003 <0.002 <0.001 Sn 0.019 0.017 --
<0.018 As 0.003 0.001 --
<0.015 Cb'" <0.002 <0.010 --
0.013 1 N 0.006 0.008 -- -- B <0.001 <0.001 -- 0.004 l- , a) Chemical Analysis by Westinghouse on irradiated Charpy specimen AT-5 removed from capsule U61, b) chemral Analysis by Westinghouse on irradiated Charpy specimen AT-64 removed frcun capsule Y. c) or Nb 4-3
TABLE 4-2 l Chemical Composition (wt7c) of the Vogtle Unit 1 Reactor Vessel Surveillance Weld Metal Weld Wire Heat no. 83653, Linde 0091 Flux. Lot No. 3536* Surveillance Wire Flux'" Actual Production
- Capsule U"" sule Yd Capsule Y" ule Y "
Elm. Program
- Test Weld knalysis Ca[malysis Cab'alysis Weld (Intermediate Analysis ,
Test Sample To Lower Shell l Weldment Girth Seam, i D 101-171) l l C 0.130 0.140 0.090 - 0.137 0.147 0.153 Mn 1.150 1.060 1.170 1.057 1.113 1.164 1.195 P 0.017 0.007 0.008 0.008 <0.014 <0.014 <0.016 l S 0.010 0.009 0.009 -- 0.0085 0.0112 0.0135 Si 0.190 0.160 0.170 -- 0.174 0.123 0.102 Ni 0.100 - 0.100 0.091 0.101 0.117 0.105 Mo 0.610 0.520 0.630 0.475 0.553 0.561 0.584 Cr 0.052 -- 0.050 0.044 0.053 0.053 0.055 Cu 0.037 0.030 0.040 0.035 0.048 0.040 0.G41 Al 0.002 -- 0.009 --
<0.019 <0.019 <0.021 Co 0.005 --
0.010 0.006 0.007 0.007 0.008 Pb <0.001 --
<0.001 -- -- -- -
W <0.010 -- 0.020 --
<0.036 <0.036 <0.039 -._
Ti 0.006 --
<0.010 --
0.011 0.011 0.012 Zr <0.002 -- 0.001 --
<0.009 <0.009 <0.010 V 0.003 0.005 0.007 0.006 0.001 0.001 0.001 Sn <0.002 --
0.003 --
<0.019 <0.019 <0.020 As 0.004 --
0.006 --
<0.015 <0.015 <0.016 Cb* <0.002 --
0.010 - 0.013 0.013 0.014 N 0.003 -- 0.021 -- -- -- -- B <0.001 --
<0.001 --
0.004 0.004 0.003 a) Surveillance weldment was made using the same wire and flux as was used for all beltline weld seams.!:1 b) Chemical analysis by Westinghouse?' c) Chemical analysis by Combustion Engineermg, Inc?) d) Chemical analysis by Westinghouse on irradiated Charpy specimen AW-12 removed from capsule U* e) Chenucal analysis by Westinghouse on irradiated Charpy specimen AW-62 removed from capsule Y. f) Chemical analysis by Westinghouse on ir adiated Charpy specimen AW-65 removed from capsule Y. g) Chemical analysis by Westinghouse on irradiated Charpy specimen AW-66 removed from capsule Y. h) or Nb ~ l l 4-4 l
i TABLE 4-3 ; Heat Treatment of the Vogtle Unit 1 Reactor Vessel Surveillance Materialst2) !, Material Temperature ('F) Time (hr) Coolant Surveillance Austenitizing: , Program 1600 25 4 Water quenched Test Plate
. B8805-3 Tempered:
- 1225 25 4 Air cooled
- Stress Relieff*)
, 1150 50 17.5 Furnace cooled l Weldment Stress Relieff') 1150 + 50 12.75 Furnace cooled a) The stress relief heat treatment received by the surveillance test plate and weldment have been simulated. 4 E a
=
4 l l 4-5
_ __ ._ _ _ _ . m - _ ._ TABLE 4-4 Chemistry Results from the Low Alloy Steel NIST Certified Reference Standards Concentration in Weight Percent NIST 361 NIST 362 NIST 363 Metals Certified Measured Certified Measured Certified Measured j Fe 95.600 95.793 95.300 95.758 94.400 94.577 . -i Co 0.032 0.030 0.300 0.297 0.048- 0.044 i
~
Cr 0.694 0.664 0.300 0.291 1.310 1.262 Cu 0.M2 0.(41 0.500 0.487 0.100 0.094 Mn 0.660 0.625 1.040 1.025 1.500 1.447 ! Mo 0.190 0.186 0.068 0.061 0.028 0.021 Ni 2.000 1.844 0.590 0.543 0.300 0.282 P 0.014 0.017 0.041 0.041 0.029 0.031 Ti 0.020 0.016 0.084 0.024 0.050 0.045 V 0.011 0.006 0.040 0.037 0.310 0.301 A1 0.021 0.019 0.095 0.080 0.240 0.234 As 0.017 0.018 0.092 0.082 0.010 <0.015 B 0.000 <0.002 0.003 0.004 0.001 <0.002 Nb 0.022 0.021 0.290 0.016 0.049 0.042 Sn 0.010 <0.019 0.016 <0.015 0.1M 0.102 W 0.017 0.047 0.200 0.202 0.046 <0.037 Zr 0.009 0.010 0.190 <0.187 0.M9 0.038 , C 0.383 0.383 0.160 ' O.161 0.620 NA S 0.0140 NA 0.0360 0.0358 0.0068 NA Si 0.222 0.208 0.390 0.414 0.740 NA NA represents elements not requested for analysis G
i l TABLE 4-4 cont. Chemistry Results from the Low Alloy Steel NIST Certified Reference Standards Concentration in Weight Percent NIST 364 NIST 4H NIST 5J Metals Certified Measured Certified Measured Certified Measured Fe 96.700 %.937 94.680 94.589 93.030 93.000 Co 0.150 0.146 -- 0.007 -- 0.009 Cr 0.%3 0.062 0.117 0.113 0.021 0.021 Cu 0.249 0.238 0.243 0.243 0.990 0.943 l Mn 0.255 0.237 0.840 0.829 0.700 0.672 1 Mo 0.490 0.468 0.017 0.015 0.005 <0.005 Ni 0.144 0.128 0.065 0.065 0.018 0.019 P 0.010 <0.014 0.124 0.143 0.241 0.244 Ti 0.240 0.262 0.024 0.030 0.044 0.057 I V 0.105 0.102 0.011 0.006 0.012 0.009 l' Al 0.008 <0.019 -- <0.020 -- <0.019 As 0.052 0.037 0.015 <0.016 0.026 0.019 B 0.011 0.011 --
<0.002 --
<0.003 Nb 0.157 0.034 --
0.014 -- 0.014 Sn 0.008 <0.019 --
<0.019 --
<0.018 l
. W 0.100 0.036 --
<0.037 -- <0.035 Zr 0.068 <0.059 --
<0.002 --
<0.002 C 0.870 NA 2.440 NA 2.370 NA S 0.0250 0.0242 0.0700 NA 0.1000 NA Si 0.%5 NA 1.34 NA 2.44 NA NA represents elements not requested for analysis I 4-7 l
l
0' REACTOR VESSEL CORE BARREL NEUTRON PAD (301.5') Z CAPSULE U (58.5')
,,. -**- 58.5 *
> l 58.5* % L
- 6_
270* g. l y (241 ') ] X W (121.5') (238.5 *) . REACTOR VESSEL 180* .' PLAN V!EW ,
; VESSEL g [ WALL l k CAPSULE ASSEMBLY CORE ; s/
'lillllllllli. f.(
. CORE MIDPLANE
- f. l 3l { N d l l N NEUTRON PAD i :s k '
/
~
CORE BARREL FLEVATION VIEW
^I Figure 4-1. Arrangement of Surveillance Capsules in the Vogtle Unit 1 Reactor Vessel r2; 4-8
- os -,.
4 a l LEGEND: AL- - INTERMEDIATE SHELL PLATE B8805-3 (LONGITUDINAL) AT - INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE) AW - WELD METAL AH - HEAT-AFFECTED-ZONE MATERIAL
\
l l 1
- 1
- - - -.m - ,
GAE "
* * * ^= *= ==$
Y = - , . . , , ; g -
=
Y =
== mn = = 7 ., ,, , y m a
- c. -
-l ll ll ,L - ,,_ .,,,c. ,,_Jj '
UUU " nnp - 4 .. c. ic., h
= , , , inu ..
-t to Tcr or vtssa. conpiC v~
i
# t i
f 1 i x 1 l I i
~ $
ANSTEC ! APERTURE g CARD I ;
~
au . . Also Avallable on Aperture Card AJ L:w 1 l l l l l l l i s M SeenNTame resuma8 X Cm CNMPTS 88tMIPT8 5 , _ _ .O m teampys egare especTe femmas AMEl AL15 AT75 ALM, Af72 AL72 f ATes ALGB A100 Alas Af85 Em AT16 r A, __ . _ _
... .m o m, .n u - Am o
u.
- Am ,,. An. Am Am ;
A,.i e Am or Am a Am o ou o _ Am a Am
.. j iTFL i. .
___ a, .. c. c., - A J. .. II .. il .L - 4 , _ . . . c. JU " tru U i }} : i
- 4, .ssa c. (col
))}e s m. - - as .isa c. (cel l 11 1 l
Ni m ,,,,, - f m
- leelr i Na '
l, s
- i l j.
2 i i i Ecw w vesssi. .. 10 00T104 W 6 Figure 4-2. Capsule Y Diagram Showing the p location of Specimens, Thermal Monitors, and Dosimeters m ( o r [- 4-9
~
3
SECTION 5.0 1 TESTING OF SPECIMENS FROM CAPSULE Y ' 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metalographic Facility at the Westinghouse Science and Technology Center. Testing was performed in accordance with 10CFR50, Appendices G and Hm, ASTM Specification E185-82m and Westinghouse Procedure MHL 8402, Revision 2 as modified by I Westinghouse RMF Procedures 8102, Revision 1, and 8103, Revision 1. I Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-110llm, j No discrepancies were found. Examination of the two low-melting point 579 F (304 C) and 590 F (310 C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579*F (304 C). The Charpy impact tests were performed per ASTM Specification E23-92Nand RMF Procedure 8103, Revision 1, on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 8301 instrumentation system, feeding information into an IBM XT Computer. With this system, load-time and energy-time signals can be recorded in addition to the 1 i standard measurement of Charpy energy (Eo). From the load-time curve, the load of general yielding l (Pay), the time to general yielding (tay), the maximum load (Pu), and the time to maximum load (tu) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was 1 observed. The load at which fast fracture was initiated is identified as the fast fracture load (Py), and the load at which fast fracture terminated is identified as the arrest load (P3 ). The energy at maximum load (Eu) was determined by comparing the energy-time record and the load-time record. The energy i at maximum load is approximately equivalent to the energy required to initiate a crack in the
- specimen. Therefore, the propagation energy for the crack (E,) is the difference between the total energy to fracture (Eo) and the energy at maximum load (Eu). ! 5-1 4
The yield stress (cy) was calculated from the three-point bend formula having the following expression: cy = (Pay
- L) / [B * (W - a)2
- C] (1) where: L = distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W = height of the specimen, measured perpendicularly to the notch a = notch depth
]
The constant C is dependent on the notch flank angle (4), notch root radius (p) and the type of loading (ie. pure bending or three-point bending). In three-point bending, for a Charpy specimen in which 4 = 45' and p = 0.010", Equation 1 is valid with C = 1.21. Therefore, (for L = 4W), oy = (Pay
- L) / [B * (W - a)2
- 1.21] = (3.3
- Pay
- W) / [B * (W - a) ] (2)
For the Charpy specimen, B = 0.394", W = 0.394" and a = 0.079" Equation 2 then reduces to: oy = 33.3
- Pay (3) where oy is in units of psi and Pay is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three point bend formula.
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-92M The lateral expansion was measured using a dial l gage rig similar to that shown in the same specification. i l Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-91 01 and E21-79(1988)ou, and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test. 52 l l i l
__ _ __ _ - _ ._-._ _ _ - . _ _ _ - __ _ _ _ __ _ _ _ _ _ ~ 1 1 Extension measurements were made with a linear variable displacement transducer extensometer. The i j extensometer knife edges were spring-loaded to the specimen and operated through specimen failure.
- The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-l 92n21, i
i Elevated test temperatures were obtained with a three zone electric resistance split-tube fumace with a I i 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a i ' thermocouple directly to the specimen, the following procedure was used to monitor specunen temperatures. Chromel-alumel thermocouples were positioned at center and each end of the gage section of a dummy specimen in each grip. In the test configuration, with a slight load on the J j specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range from room temperature to 550'F (288'C). During the actual testing, the grip l ] temperatures were used to obtain desired specimen temperatures. Experiments indicated that this method is accurate to 2*F. 1 3 The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined l directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength i . j-were calculated using the original cross-sectional area. The final diameter and final gage length were ] determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true l stress at fracture) and percent reduction in area was computed using the final diameter measurement. - s I i l 5.2 Charpy V-Notch Impact Test Results f 1 The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule Y, which was irradiated to 1.24 x 10" n/cm2(E > 1.0 MeV), are presented in Tables 5-1 through 5-8 and are compared with unirradiated results"1 as shown in Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule Y materials are ) summarized in Table 5-9. I 1 j l ] l j Irradiation of the reactor vessel intermediate shell plate B8805-3 Charpy specimens oriented with the I i
, longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal j orientation) to 1.24 x 10" n/cm2(E > 1.0 MeV) at 550 F (Figure 5-1) resulted in a 30 ft-lb transition j- temperature increase of 40 F and a 50 ft-lb transition temperature increase of 55 F. This results in an
] irradiated 30 ft-lb transition temperature of 25 F and an irradiated 50 ft lb transition temperature of l 53 i i - _ _ _~ , _
55 F (longitudinal orientation). The average upper shelf energy (USE) of the intermediate shell plate B8805-3 Charpy specimens . (longitudinal orientation) resulted in no energy decrease after irradiation to 1.24 x 10 n/cm2 (E > 1.0 MeV) at 550 F. This results in an inadiated average USE of 122 ft-lbs (Figure 5-1). hradiation of the reactor vessel intermediate shell plate B8805-3 Charpy specimens oriented with the
~
longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation) to 1.24 x 10 ' n/cm2(E > 1.0 MeV) at 550 F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 20 F and a 50 ft-lb transition temperature increase of 30 F. This results in an irradiated 30 ft-lb transition temperature of 35 F and an irradiated 50 ft-lb transition temperature of 95 F (transverse orientation). The average USE of the intermediate shell plate B8805-3 Charpy specimens (transverse orientation) 2 resulted in no energy decrease after irradiation to 1.24 x 10 n/cm (E > 1.0 MeV) at 550, F. This results in an irradiated average USE of % ft-lbs (Figure 5-2). 2 Irradiation of the surveillance weld metal Charpy specimens to 1.24 x 10 ' n/cm (E > 1.0 MeV) at 550 F (Figure 5 3) resulted in no 30 ft-lb and 50 ft-lb transition temperature increases. This results in an irradiated 30 ft-lb transition temperature of -40 F and an irradiated 50 ft-lb transition temperature of 25 F. The average USE of the surveillance weld metal resulted in an energy decrease of I ft-lb after 2 irradiation to 1.24 x 10 n/cm (E > 1.0 MeV) at 550 F. This results in an irradiated average USE of 144 ft-lbs (Figure 5 3). Irradiation of the reactor vessel weld HAZ metal Chagy specimens to 1.24 x 10 n/cm2 (E > 1.0 MeV) at 550 F (Figure 5-4) resulted in 30 ft-lb and 50 ft-lb transition temperature increases of 25 F. This results in an inadiated 30 ft-lb transition temperature of -50'F and an irradiated 50 ft-lb transition temperature of -25 F. The average USE of the weld HAZ metal resulted in an energy decrease of 12 ft-lbs after irradiation to 1.24 x 10 n/cm 2(E > 1.0 MeV) at 550 F. This results in an irradiated average USE of 124 ft-lbs - (Figure 5-4). 5-4
i l l l l The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test l temperature. l
~
A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the l various Vogtle Unit 1 surveillance matedals with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2D1 is presented in Table 5-10 and led to the following conclusions: o The 30 ft-lb transition temperature increases for the capsule Y surveillance materials are less than the Regulatory Guide 1.99, Revision 2 predictions. o The Upper Shelf Energy decreases of the capsule Y surveillance materials are less than the Regulatory Guide 1.99, Revision 2 predictions. 1 i The load-time records for the Charpy tests are provided in Appendix A. l 5.3 Tension Test Results The results of the tension tests performed on the various materials contained in capsule Y irradiated to 2 1.24 x 10" n/cm (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated l resultst2i as shown in Figures 5-9 through 5-11. l The results of the tension tests performed on the intermediate shell plate B8803-3 (longitudinal l 2 l orientation) indicated that irradiation to 1.24 x 10" n/cm (E > 1.0 MeV) at 550 F caused a 0 to 4 ksi l increase in the 0.2 percent offset yield strength and a 0 to 6 ksi increase in the ultimate tensile strength when compared to unirradiated datar21 (F gure 5-9). The results of the tension tests performed on the intermediate shell plate B8805-3 (transverse 2 orientation) indicated that irradiation to 1.24 x 10" n/cm (E > 1.0 MeV) at 550 F caused a 0 to 3 ksi l increase in the 0.2 percent offset yield strength and a 0 to 3 ksi increase in the ultimate tensile strength I, when compared to unirradiated datat21 (Figure 5-10). The results of the tension tests performed on the surveillance weld metal indicated that irradiation to 1.24 x 10" n/cm 2(E > 1.0 MeV) at 550 F caused a 1 to 4 ksi increase in the 0.2 percent offset yield 5-5 l l
strength and a 2 to 5 ksi increase in the ultimate tensile strength when compared to unitradiated datam (Figure 511). The fractured tension specimens for the intermediate shell plate B8805 3 material are shown in Figures 512 and 5-13, while the fractured specimens for the surveillance weld metal are shown in Figure 5-14. The engineering stress-strain curves for the tension tests are shown in Figures 5-15 through 5-20. l 5.4 Compact Tension Specimen Tests l Per the surveillance capsule testing contract, the 1/2T compact tension (CT) specimens were not tested. The 1/2T CT specimens are being stored at the Westinghouse Science and Technology Center Hot Cell facility. e 1 l I l 5-6 i j
i i TABLE 5-1 3~ Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 2
- Irradiated at 550 F to a Fluence of 1.24X10* n/cm (E > 1.0 MeV)
(Imagitudinal Orientation) I~ Sample Temperature Impact Energy Lateral Expansion Shear Number (*F) ( C) (ft-lb) (J) (mils) (mm) (%) i A147 -25 -32 6 8 5 0.13 10 l AL69 0 -18 23 31 18 0.46 - 15 1 AL74 25 -4 17 23 17 0.43. 20 AL61 25 -4 23 31 20 0.51 20 l
- AL65 35 2 53 72 39 0.99 35 i AL64 50 10 57 77 44 1.12 40 i
i AL62 60 16 41 56 31 0.79 30 3 ! AL71 75 24 82 111 51 1.30 60 1 1 AL73 100 38 74 100 56 1.42 70 AL72 125 52 79 107 60 1.52 80 ) AL75 150 66 % 130 62 1.57 90 $ A1A6 175 79 94 127 69 1.75 95 o AII>8 225 107 132 179 78 1.98 100 l A143 275 135 130 176 80 2.03 100 ) AL70 300 149 133 180 88 2.24 100 i l 1 5-7 i
l TABIE 5-2 Charpy V notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 2 Irradiated at 550'F to a Fluence of 1.24X10 n/cm (E > 1.0 MeV) (Transverse Orientation) ,
~
Sample Temperature Impact Energy Lateral Expansion Shear Number ( F) ( C) (ft-lb) (J) (mils) (mm) (%) j l AT65 -50 -46 17 23 12 0.30 5
. 1 AT64 -25 -32 28 38 20 0.51 15 l
AT66 -20 -29 21 28 15 0.38 10 - AT72 0 -18 24 33 20 0.51 10 l AT75 25 -4 20 27 18 0.46 15 l AT70 35 2 33 45 26 0.66 15 ! AT68 75 24 45 61 34 0.86 45 I I AT63 100 38 47 64 40 1.02 45 , AT73 110 43 55 75 48 1.22 70
- AT69 125 52 65 88 50 1.27 70 I AT71 150 66 67 91 53 1.35 75 AT74 175 79 81 110 66- 1.68 90 .
AT67 225 107 100 136 73 1.85 100 AT61 275 135 109 148 66 1.68 100 AT62 300 149 109 148 75 1.91 '100
=
e f l 9 1 l 5-8 a
(. _. . - - .. . ~ _ _ - - . . . - - . I TABLE 5-3 Charpy V-notch Data for the Vogtle Unit 1 Surveillance Weld Metal 2 Irradiated at 550 F to a Fluence of 1.24X10P n/cm (E > 1.0 MeV) Sample Temperature Impact Energy Lateral Expansion Shear Number ('F) ( C) (ft-lb) (D (mils) (mm) (%) AW66 -50 -46 13 18- 11 0.28 10 AW68 -37 17 23 14 0.36 15 -
- AW65 -25 .-32 15 20 14 0.36 20
~ AW74 -25 -32 67 91 52 1.32 50 AW62 -10 -23 98 133 67 1.70 70 AW67 -5 -21 80 108 57 1.45 60 AW61 0 -18 94 127 68 1.73 70 I
l AW63 5 -15 114 155 71 1.80 80
- i. .
AW70 25 -4 70 95 52 .1.32 55 AW72 40 4 109 148 72 1.83 80 AW71 60 16 112 152 74 1.88 80 l AW69 100 38 124 168 86 2.18 95 l AW75 175 79 134 182 87 2.21 100 AW73 275 135 153 207 76 1.93 100 l AW64 - 300 149 145 197 81 2.06 100 a 5-9
TABLE 5-4 Charpy V-notch Data for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) Metal Irradiated 2 at 550 F to a Fluence of 1.24X10* n/cm (E > 1.0 MeV) Sample Temperature Impact Energy Lateral Expansion Shear , Number (*F) ( C) (ft-lb) (J) (mils) (mm) (%) AH75 -100 -73 19 26 14 0.36 5 i AH68 -50 -46 28 38 21 0.53 15 AH63 -35 -37 15 20 22 0.56 20 AH67 -30 -34 75 102 49 1.24 55 , AH72 -25 -32 78 106 52 1.32 60 AH66 -25 -32 89 121 55 1.40 65 AH71 0 -18 .65 88' 43 1.09 50 AH74 25 -4 121 164 69 1.75 85 AH65 45 7 114 155 69 1.75 85 AH61 60 16 78 106 51 1.30 65 AH73 100 38 134 182 77 1.% 100 , AH69 150 66 109 148 71 1.80 100 AH62 200 93 114 155 71 1.80 100 AH70 250 121 141 191 77 1.% 100 AH64 300 149 123 167 75 1.91 100 5-10 1 I
. . +
TABLE 5-5 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated at 550 F to a Fluence of 1.24 X 10 n/cm2 (E > 1.0 MeV) (Iengitudinal Orientation) Normalized Energies 2 (ft-lb/in ) Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Imad Yield Load Max. Load IAad Stress Stress Sample Temp. E,/A E,/A Pay toy Pu tu Pg PA OY (NSi) Eo Er/A No. ( F) (ft-lb) (lbs) (psec) (Ibs) (psec) (lbs) (Ibs) (ksi) AL67 -25 6 48 25 24 2817 0.12 3043 0.14 3013 0 94 97 AL69 0 23 185 141 44 3408 0.16 4074 037 4067 169 113 124 AL74 25 17 137 67 70 3292 0.16 3495 0.23 3479 318 109 113 AL61 25 23 185 130 55 3327 0.17 3950 0.36 3944 229 111 121 AL65 35 53 427 304 123 3316 0.16 4301 0.68 4246 158 110 126 AL64 50 57 459 297 162 3155 0.15 4217 0.67 4081 526 105 122 AL62 60 41 330 241 89 3112 0.15 4138 0.57 4135 784 103 120 AL71 75 82 660 298 363 3234 0.16 4233 0.67 3706 456 107 124 AL73 100 74 596 294 302 3187 0.17 4195 0.69 3820 1708 106 123 AL72 125 79 636 281 355 2894 0.14 4067 0.67 3584 2014 % 116 AL75 150 % 773 351 422 2840 0.14 4059 0.82 3455 1938 94 115 AL66 175 94 757 279 477 2802 0.15 3970 0.69 3021 1791 93 112 AL68 225 132 1063 284 778 2766 0.14 4054 0.68 ** ** 92 113 AL63 275 130 1047 278 769 2742 0.14 3954 0.68 ** ** 91 111 AL70 300 133 1071 l 262 809 2634 0.15 3797 0.67 87 107
** Fully ductile fracture.
TABLE 5-6 Instrumented Charpy impact Test Results for the Vogtle 2 Unit 1 Intermediate Shell Plate B8805-3 Irradiated at 550 F to a Fluence of 1.24 X 10 n/cm (E > 1.0 hieV) (Transverse Orientation) Normalized Energies 2 (ft-lb/in ) Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Chagy Max. Prop. Load Yield Load Max. Imad Load Stress Stress Sample Tem En En/A eda Ep/A Por toy P tu Pg P3 oy (ksi) No. (F (ft-Ib) (lbs) (psec) (Ib5) (psec) (Ibs) (lbs) (ksi) AT65 -50 17 137 100 37 3738 0.17 4070 0.28 4054 0 124 130 AT64 -25 28 225 200 25 3677 0.17 4354 0.47 4335 0 122 133 AT66 -20 21 169 129 40 3553 0.17 4066 0.34 4040 0 118 127 AT72* O 24 193 - - - - - - - - - - AT75 25 20 161 93 68 3319 0.16 370% 0.28 3685 446 110 117 y AT70 35 33 266 206 60 3425 0.17 4250 0.49 4237 181 114 127 AT68 75 45 362 244 118 3268 0.16 4237 0.57 4233 680 109 125 AT63 100 47 378 235 144 3070 0.15 4043 0.57 4008 1573 102 118 AT73 110 55 443 231 212 3147 0.16 4 040 0.56 4002 1545 105 119 AT69 125 65 523 235 288 3147 0.16 4078 0.57 3721 1865 105 120 AT71 150 67 540 262 278 2895 0.15 4027 0.63 3720 2290 % 115 AT74 175 81 652 268 384 2892 0.14 3972 0.65 3150 1950 % 114 AT67 225 100 805 282 523 2829 0.14 4056 0.67 ** ** 94 114 AT61 275 109 878 269 609 2724 0.14 3860 0.67 ** ** 90 109 AT62 300 109 878 272 606 2811 0.17 3884 0.69 ** ** 93 111
- Data not available due to machine malfunction.
** Fully ductile fracture.
8 4 , D h
1 TABLE 5-7 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Surveillance Weld Metal 2 Irradiated at 550 F to a Fluence of 1.24 X 10 n/cm (E > 1.0 MeV) Normalized Energies (ft-lb/in') Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Load Yield lead Max. IAad Load Stress Stress Sample Temp. En Eo/A E,/A E,/A Pyo ty o Pu tu P, PA UY (kSi) No. ( F) (ft-lb) (lbs) (psec) (lbs) (psec) (Ibs) (Ibs) (ksi) AW66 -50 13 105 48 57 3815 0.16 3958 0.18 3935 314 127 129 AW68 -35 17 137 67 70 3811 0.17 3885 0.22 3879 519 127 128 AW65 -25 15 121 46 75 3605 0.16 3727 0.18 3705 754 120 122 AW74 -25 67 540 322 217 3737 0.17 4442 0.68 4229 1030 124 136 y AW62 -10 98 789 225 564 3562 0.17 4244 0.52 3160 1535 118 130 U AW67 -5 80 644 313 331 3676 0.17 4410 0.67 4022 2095 122 134 AW61 0 94 757 315 442 3554 0.16 4318 0.69 3892 2377 118 131 AW63 5 114 918 310 608 3499 0.18 4223 0.69 2837 1512 116 128 AW70 25 70 564 310 253 3503 0.16 4310 0.68 4297 2571 116 130 AW72 40 109 878 309 568 3399 0.16 4265 0.69 3331 2013 113 127 AW71 60 112 902 304 597 3366 0.16 4208 0.69 3389 2459 112 126 AW69 100 124 998 294 704 3306 0.17 4125 0.68 3112 2451 110 123 AW75 175 134 1079 357 722 2965 0.15 4043 0.82 ** ** 98 116 AW73 275 153 1232 341 891 2716 0.14 3898 0.81 ** ** 90 110 AW64 300 145 1168 266 902 2701 0.15 3747 0.69 ** ** 90 107 ,
** Fully ductile fracture.
TABLE 5-8 Instrumented Charpy Impact Test Results for the Vogtle Unit i Surveillance IIeat-Affected-Zone (IIAZ) Metal 2 Irradiated at 550 F to a Fluence of 1.24 X 10 n/cm (E > 1.0 MeV) Normalized Energies (ft-lb/in') Tune 'Ilme Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Load Yield Load Max. Imad Imad Stress Stress Sample Temp. En Eo/A EJA E,/A Pay ty o P,, t, i Pg P, or (ksi) No. ( F) (ft-lb) (Ibs) (psec) (Ibs) (psec) Obs) (Ibs) (ksi) AII75 -100 19 153 91 62 4195 0.17 42 % 0.25 4283 314 139 141 AH68 -50 28 225 110 115 3916 0.17 4152 030 4149 1650 130 134 AII63 -35 15 121 110 11 3737 0.17 4137 030 4133 1500 124 131 AII67 -30 75 604 231 373 3709 0.16 4423 0.51 3763 979 123 135 AII72 -25 78 628 321 307 3682 0.17 4476 0.69 4409 3147 122 135 AII66 -25 89 717 324 393 3732 0.17 4515 0.68 3761 1094 124 137 AII71 0 65 523 291 233 3689 0.17 4431 0.63 4163 1268 123 135 AII74 25 121 974 316 659 3481 0.16 4352 0.69 3054 1818 116 130 AH65 45 114 918 300 618 3337 0.16 4210 0.68 2535 1499 111 125 AH61 60 78 628 218 410 3369 0.16 4160 0.52 3321 20(M i12 125 3257 0.17 4202 ** ** AII73 100 134 1079 331 748 0.75 108 124 AII69 150 109 878 288 590 3315 0.17 4089 0.68 ** ** 104 120 AH62 200 114 918 277 641 2883 0.16 3957 0.68 ** ** % 114 AII70 250 141 1135 351 784 2980 0.17 4049 0.82 99 117 716 0.16 ** ** 94 AH64 300 123 990 274 2839 3885 0.68 112
** Fully ductile fracture.
8 9 3 4 % 4
TABLE 5-9 2 Effect of 550 F Irradiation to 1.24 X 10 n/cm (E > 1.0 MeV) on the Notch Toughness Properties of the Vogtle Unit 1 Recctor Vessel Surveillance Materials Average 30 (ft-lb) (*' Average 35 mil Lateral (') Average 50 ft-lb (*) Average Energy Absorption (*) Transition Temperature ( F) Expansion Temperature ( F) Transition Temperature ( F) at Full Shear (ft-lb) Material Unirradiated Irradiated AT Unirradiated Irradiated AT Unitradiated Irradiated AT Unirradiated Irradiated AE i Plate B8805-3 - 15 25 49 10 45 35 20 55 35 122 132 + 10 (longitudinal) Plate B8805-3 35 20 55 75 20 65 95 30 96 106 + 10
? 15 G (transverse)
Weld Metal - 40 - 40 0 - 35 - 25 10 - 25 - 25 0 145 144 -1 IIAZ Metal - 75 -50 25 - 45 - 45 0 - 50 - 25 25 136 124 - 12 , (a) " Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-1 through 5-4) f
Y l j TABLE 5-10 Comparison of the Vogtle Unit i Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 30 ft-lb Transition _ Upper Shelf Energy Temperature Shift Decrease M
- Material Capsule. (X 0 ' n/ 2) . f*
It M ict y Me Intermediate Shell U 0344 27 15 15 0 Plate B8805-3 y (longitudinal) Y 1.24 41 40 20 0 Intermediate Shell . U 0344 27 0 15 0 Plate B8805-3 (transverse) Y 1.24 41 20 20 0 Weld Metal U 0344 - 23 15 15 0 Y 1.24 35 -0 20 -I H A Z M etal U 0344 - 0 -
-6 Y 1.24 -
25 -
-9 j (a) Based on Regulatory Guide 1.99, Revision 2 methodology using Mean wt. % values of Cu and Ni.
s e
# g . 8 g % g m._m_m. _ . _ _ ._____m ___ _ _ . _ _ . .___.___..m___._._ _ _ _ _ _m___m_.____. . , -. m m__- .- ~ , - _ - ~ _ _ ,_._c-__________
.______.__-m.___.___m_______ _ __ _ _ _ _ _
=
TABLE 5-11 2 Tensile Properties of the Vogtle Unit 1 Rerctor Vessel Surveillance Materials Irradiated at 550 F to 1.24 X 10' nkm (E > 1.0 MeV) Test 0.2% Ultimate Fracture Fracture Fracture . Uniform Total Reduction Sample Tem Yield Strength Load Stress Strength Elongation Elongation in Area Material Number p. Strength - (ksi) (kip) (ksi) (ksi) (%) (%) (%) (*F) (ksi) 1 Plate B8805-3 AL13 85 72.8 96.8 3.00 1823 - 61.1 10.5 '24.9 66 (longitudinal) Plate B8805-3 ALl4 200 703 88.6 2.80 .153.0 57.0 9.9 22.5 63 - (longitudinal) Plate B8805-3 ALIS 550 64.2 90.1 3.25 137.2 66.2 10.5 20.9 52
% (Longitudinal Plate B8805-3 AT13 100 703 953 3.20 163.4 65.2 10.3 23.2 60 ;
(transverse) Plate B8805-3 AT14 225 68.8 88.6 3.00 169.8 61.I' 9.9 21.8 64 (transverse) Plate B8805-3 ATIS 550 67.2 90.7 334 149.7 67.9 10.5 .19.4 55
- (transverse)
- Weld Metal AW13 175 72.8 85.6 2.50- 170.0 50.9 9.0 '23.4 70 Weld Metal AW14 25 74.9 913. '2.65 180.2 54.0 11.1 30.0 70 Weld Metal AW15 550 - 67.5 85.6 2.60 158.0 53.0 9.6 23.1 66 s
i 1 4 i _ _. , , . _ __- . . . - _
i l l ( C)
-150 -100 -50 0 50 100 150 200 250 100 1 I I I Ias ~ Ij;;i2 j
I I 2s 8 80 2 w 60 - o a Si 40 - e s s g 20 0 I iSI 2 i i i i 1 . 100 2.5 l n - 9 e</' ym 80 oo - - 2.0 o- 60 - * - 1.5 - x c 1.0 6 l 40 - g 0.5
$20 3 I I I I I I I 0 O
! 160 200 140 o y ve % 120
- e v -
160 o o 8
- 9 100 -
o - U 0 * - 120 m 5 80 - e O Gi B 60 - - 80 5 n y 40 - * (r - 40 20 - ** 0 0
-200 -100 0 100 200 300 400 500 TEMPERATURE ( F)
O IMRRADIATO 19 e 5tRABIATD CiS(ff), RUDCE 1.24 x 102 6 E ) U W Figure 5-1 Charpy V-Notch Impact Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (IAngitudinal Orientation) 5-18
1 l l l ( C)
-150 -100 -50 0 50 100 150 200 250 I I I I Ia - lj lj l I 100 - - - - -
8 80 s 2
%! 60 -
W w 40 - o 20 I I I I I I 0 100 2.5 y- 80 - o ,,,;; OR - 2.0 0 e ex 60 - - 1.5 c-
- - 0
'] 40 Yr 1.0 5 20 - -
0.5 1 I I I i i 1 0 0 120 160 em 100 - oM ,
~
120 g 80 - T - C
- 9
- 60 - -
80 5 40 - 2fr - 40 20 - o o l I I I I I i 0 0
- 200 -100 0 100 200 300 400 500 TEMPERATURE ('F) l
)
o LMRRANAD , 19 2 j e IRRAMAD G50'R ruDCE L24 x 2 n/cm E > 18 MeV) l l Figure 5-2 Charpy V-Notch Impact Properties for Vogtle Unit 1 Rt _ tor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) 5-19 i t
1 ('C)
-150 -100 -50 0 50 100 150 200 250 I l I 100 I l i 0
If f l f ^ ^I f , 3\
-
- o
@ 80 y ,
SE 60 - E o
" 40 -
o i 2- l l l l l 0 100 p j , 2.5 g; 80 q.o** w - .o g eo 2.0 a- 60 1.5 2 m d 40 - o/ - 1.0 v E m*r 320 2s - 0.5 0 t a - 21 1 1 1 1 0 200 180 - 240 160 - o g , . a -m - 200
,140 -
,- o ~T 2
.o O 160 -
!.120 ,
D 100 - 0-2 3 120 g860 o. o - 80 40 - _) _ o*r . l l l l l l 0
-200 -100 0 100 200 300 400 500 TEMPERATURE ( D o umAMATO 4 e RRAMATED c50*D, F11DCE 124 x1910 n/cn 2 g 3 gj y,y) l l
Figure 5-3 Charpy V-Notch Impact Properties for Vogtle Unit 1 Reactor Vessel Surveillance Weld Metal 5-20 l l
(*C)
-150 -100 -50 0 50 100 150 200 250 I I I I gl 3 1 I I 100
- o ^0 0 0{0 0 8 80 oo I
o\
% 60 -
" 40 -
oo 20 e' 0
- knk* I I I i 1 100 2.5
^ @
3 80 -
.o v . . -
2.0 6 o o w w - e 60 1.5 x o e 2 9e 9 z'fr 40 - 1.0 0.5 320 2 s I- I I I I I 0 0 160 o 2 200 140
/ 2'o * . 9 e '
120 o
- 160
^ *o e ~
y100 g g g e - 120 g v
" 80 -
e ni
@ 60 80 0 -
dr 40 - 2rr - 40 20 , 0 0
-200 -100 0 100 200 300 400 500 i TEMPERATURE ( F)
O LMRRAMATO e IRRAMATO CISID, rWDCE 124 x 1[n/cn2 g > 3,3 g,y) Figure 5-4 Charpy V-Notch Impact Properties for Vogtle Unit 1 Reactor Vessel Weld Heat-Affected-Zone Metal 5-21
y mn ,a ms_. .
=:4, .. v . .
..->mm ,
~
i .,,4 4 _c"w 4; c tenzsw n ... J ,.f .d -
?-em ff x%i
.a m & lT .,
.g
?h,.
fSf.Cl 9nea v hh,f,, -~ r
- {
't - ,ti s ,
HWdh: N$kd im, rw3 *' w<
- w. ewe
-Qv m .g g a
.M,.%g l pg - ty/WQ; p e shyg ,. g 9m :f 9 N.g;
. l Q.
' ~
e Oq}}?Q ' '+ "M]
. ikfR;( qt ..= "
> - . . .a . w - .. . - -- . a.- n AL67 AL69 AL74 AL61 AL65
- g: . .
u ..m. . -4r.. _ s tW e m a, sgaRyp 7: lyr t:- 4g *t ,p
;;p,.,. .. .y' 1 " lyf.
9:14 r gh u3 q*Q+ ~ 3 g: u; . 4(p'kl r'Q, .- h[ p +;;p;, "' qi;,
- 3. f -",,
wM j
'; . = ,. .Q . ,
**6 -.+ 5 *
[ g.e m
.A .
- r. y & $@D ' yy v:
%SN4 e.
- 3 Q;Q
.er-xm ..3 - un,y .my ; 29 , .
+ w;rp,m}
n , ;yg. ;. . ryx , m. -
, :;p ,
AL64 AL62 AL71 AL73 AL72 ,
-.,p.
, . w: m y -~s- . . . , - , . . . . . , . . . ,7.,,
w m;y7; : pg.-q ,
. . .y9, +.*,,
2.;.: 99 :*= ,
.{ , . .A y,L.1%&y:~y j,. g ;
;W m s;w : ,7 >> 4
. :. d, ..e.
-f}p% ' :S@jyf ~
' Tkg .
%% j .c
%~lA{)
w m m w:, n ..
?%yirt %5l '
QQ:{ , E,.c. u.aO j.~~-fav.- R .. - 1
,--.3 1 ,
g._ ,- . s, w- e ag.yp
- .3 - .:a >
AL75 AL66 AL68 AL63 AL70 Figure 5-5 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor - Vessel Intermedia *e Shell Plate B8805-3 (Longitudinal Orientation) 5-22
l l e
%am;y-; %,..:.,;.a q . ' ' .mg3_.
rdv+v : ' v g:qm g,3
- .c : 9knn-wg =rj
&yQq;; vm .
^ jzQ' . weA et +g et~
)'IN.. AM jNi u NM@f% $ M s!" $yfh.1, ' Q,ggskf$$. -
# q?I A #?@4 - w $
- k. gg4%g\.
y;- ,
?;mNb.~y * " ..
A
- 1#
i< -^' jQ W, "..'s
'%n,m "W :; 7's~ - ? -
f g,.-S :.33 igkg.i85C 7 MGk. . . -.,,- - jffgg WSQlTQQ%. JQ;;Q ~ , - - ..
~l .
w.. , , .- aWMbQ}. y u. 6
.. - ~ - . .
a:Meq AT65 AT64 AT66 AT72 AT75 l g >c
. ,M 5 -
w:.:: g, gg gps: 4 m n- , , g=yig p , er xwe Ms w - wgm
.w, -
.g :.
,m , . .
J , A am '- >, , ,' m_ Wpi; < yh7:T i t l a? Rgg y y yl$ h a_ x a
.m s.. .
-u.. _ . . ~ . ~; . t
- y_-- .
. h .. ~ ..M
- e. m ;.
AT70 AT68 AT63 AT73 AT69
- . ~ ~ . . , . - , . ,. .
y.s. 3 3 ... c 7,. m3 ,, y c,y;, .l eu-g,j.ggjj' ,,, ; ;,Q.p", s gjpy-g:g j 42 . wp +wm v gy gbm - gagi. g.5*4,0 ~ . 9 + ,. M'y;; ' 4~ n;yp;t.. . s
-[ .
s y n .. / d- - h" .%.'sr gs ' ^ ' * ' ' ' ' ' -'
, f 1 up ','N I
3 3 filY
-s
.; ;R e
$s; ; 'W"
c y' ap O i; ' we .
. 4,% ?m" p 'yw h .,.'
.3 3C%
W-
, ' -s QM. *p ' '
{ .. _m . m . _ --
.%4 :
- j 2e .,
,m; . g.s
.3..
i AT71 AT74 AT67 AT61 AT62 lO l ' ~ Figure 5-6 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) 5-23 _ - _. _. ~ . _
1
\
- p. n ,,,, .,: - . -, g.,., . - ,__ -,.
.y v .. - . : g~ - ., x; ver ,. s iz@@ir :
', : W . -u wo
~ i qvsti i f '
- j. m[$ 3 Y :iy *
.};
l' l $1 , , l
.. a: , n.wp;. m < '. - - > - t
[h;F , .3 7 [ < , . ,
- ' ~: R&? i ,
w m -
. . .~ . .u.. b. , . :.. _ a., . . . .
~
AW66 AW68 AW65 AW74 AW62 y_.---, . . , . . , - , . . _ - . _. _ . 7._ _ , , , . _,. _ . _ , , . . . _ . . _ . , 3h l dyT ' ' @y;ak L&&{ i
' & gl~l
;:n,.
' A %Wi py$ &ff
' VA)$:
q , 7
~,7 .] x7t a r .. ,[. . T ^^ .
l g 'l' ? ^
.=q s
> 2
-- fw
~,.6,u. _l... ..w i"w.w. . . ' ': t . me u, Sh[n' $J f,g g
i $%=rg::' . j . '
. = . = ~ . - . : __. __. -
AW67 AW61 AW63 . AW70 AW72 ,
, .. , , -..y m ..
7 :- . . 1, . . _ , ,, w _. __ .. _
- me ' _ . en - ,- m %z yQ6 M ;
+ '
}b; o:n
~OL Wy ,
yl T_ww:n i
~pgj'u-
%y;;;b .g
.,;g ~, , y -
%gf i$ %; .,
-s. .m,-
( %, ,, s,: .w:: y' wp pyg. .,w
.= ..
4
,W . ,,.
+
AW71 AW69 AW75 . AW73 AW64 1 I Figure 5-7 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Stuveillance Weld Metal 5-24
k i j a i i 1 p....
- n. >.. , .
r .m ~~;ma,c: y ,c, .. . ,, , ;q ~, .pp 1
.~~ ey ; * ,
. m ._ ,, y , m47 -
w:.m,> l ,
* ;. ? ,
k Sh [ '~~f
'h3r , , . , c.yl yy .y%n[ .:,.-
r t w
} :p s
%;+g%
Aqw o
,r;mw m e:9n F:a ,v. >
4 sp+i esp & :@, ..a < r q w
;y;
. -a ~
a ..; c, e, , y+f ;;}; t .
~ ,
, ,, ' ; &$:p f.l . i, ,
(' ( g. < -s Y] t.ma m .
._.s:n: h gg g ._
4 AH75 AH68 AH63 AH67 AH72 a l _ y~r ~mvv; " w w r j 3 n w g~ .v
! ( ]_g..p_
- me":;rgc >
5 o m m jkjw%?N ~
,Wm - e ##c. ,m A yy.m;qqqq$[q
?" ?
awh$ 1
,- g ,
/A '
n n M:qt, w, env -s.. avym gg
.-w+ , .
,, > ,. . c q.
a -
^
- E 4
,~
3: d
;:;n['7lL s
- w:r u;w : eo a .,
+ <
r. i
< 4 Mr ,
N x i $h
.. .kg a, '(.
e$s,.:pntf s l g:1 ' ukM '-
, i
' ema15W ' N 9
.t erve- ,
y; . z. , . . L fgheQ:jgjp=, UQ( ,,
- e. ~
-.:. u : n .
- a. 3- Q ki j, AH66 AH71 AH74 AH65 AH61 e
>* e+ws -r grma-. "***+ -**+tv""'7+-*W;me -
* e ~. ....m.--
+-}"*'*+f- . . . . , 7 r 4
.g. _ . . -
m_..'.,.' . ww- n . ,.
. ,em ,. -
r, g~., y.. a... c.., s -.:y a . 4
.-..m. J-=l';m + .
3Q+,. .._ y[@g%fd47 * (? j,k' eq' .:t[ ,[ Rl,
- ~ ~
- S y Qx~q' lQn% w .
y' g.. .
- i m
3.. v .
'y g~ ~; ,m 4 _ ..
a . ]c pg4 e }, dig,;g , I
$y' n$4C O y p . nMjn ; ta p, q - .x(.
4 . - _.-u. ,a .. x L = . . . .; ...,_a.. ' AH73 AH69 AH62 AH70 AH64 1 7 1 1 j- Figure 5-8 Charpy Impact Specunen Fracture Surfaces for Vogtle Unit 1 Reactor t Vessel Heat-Affected-Zone Metal i 1 1 2 5-25 i
t (*C) 0 50 100 150 200 250 300 l l I l 800 l l l-110 700 - -! 100 ULTNTE TDGILI STRDETH 2
\
^ 90 -
H 3 v s 600 e o n 80 - O E ' d = - 500 . M 70 - V + o - 60 - 02% Y1 ELD STRDEM 3 - 400 50 -
~
I ! I 40 i 80 RDUCTD IN AREA 70 - C 0 , 60 - 8 50 -
]
c g 40 D a 30 - mTAL EDEATB C i i A 20 - A e
- s@
uGTIRM ELSEATB 10
- ^ ^'
0 0 100 200 300 400 500 600 TEMPERATURE (*F) l A O tMRRADIATO A 91RRADIATO AT 550*f, FLIDCE 124 x 10 0a/cn 2(E ) 1,0 MeV) i i Figure 5-9 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) - 5 26 i _ , - , . , - - - - . . . . _ __.. . -_ _ - . . . . . - , , , . . - , _ , , . . . . _. . . . _ _ . . _ . _ _ _ . . . . . _ . _.I
('C)- 0 50 100 150 200 250 300 120 800 l l l l l l l-110
~
700 I
~
100 -2
\ (LTNTE TDGH STRDETH ,
9 90 -- i - 600 h d n 80 -
!b i M
N G" - 500 (
. u 70 -
e -
- Ae ,
60 - - 400 g 3 50 - 40 80 , 70 mucTEN N NtEA O _e n 60 - 0V o 0 50 - C 3 40 - a 30 A TOTAi. nasATIN 20 -
. 10 CO O O 0
0 100 200 300 400 500 600 TEMPERATURE ( F) A O INRRAMATD l A G RRAMATD AT 550*F, FLDCE L24 x 0 2 E )la MeV) 13 n/cn Figure 5-10 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) 5-27 I
(*C) 0 50 100 150 200 250 300 120 800 l l l l l l- , 110 700 .. 100
. S 90 ILTIMATE TEMI STilDETH ,- 600 ,
e- a s2 o n 80 - zs % . 500
'$' 70 - -
t" D . l L 60 - R2% YIED STRDETH 400 )
)
l 50 40 80 70 e
'c . g 'Ng l sucia in m e 60 -
l 8 50 - C g 40 TDTAL Od E T S 2
$ 30 - 7 20 10 *g - " " " t' a -
l I. I I i 0 0 100 200 300 400 500 600 TEMPERATURE (*F) l l A O INDIRADIAU A 91RRAMAD AT 5507, atacE 124 x 10 0 n/cn2g > to ney)
~!
Figure 5-11 Tensile Properties for Vogtle Unit 1 Reactor Vessel Surveillance Weld 1
~
Metal , 5-28 l r
l t t d i < \ ' ~
! .A, ;;Q ,g -iff h
-~; .-
sp1..M.
.,m,. 3 i -Q f k :M. w, 3 ,R 8i .j . .
, LY:".
,. ;4
~ v >;20;in, -
m ...,,..
.. . . . , g;. . . , .
l-4 m 9:!..i? _ ., , q:.i, W i,g trs $.n07'
- t. 7
- , h. .. uwk &.t @,;, p ,
- m s; w :
, w f i , *
~
p'[*L- :+%grojseum m , , , es IN. i -
. ~
x
,::sw(iM. y;y ;gG W .
4.,~- N,EV a:.. l,e.M:%ide l i:j.1 % .nr.L W.L-M M hM ih'.~4-~:.W .- Abd } 5 Specimen AL13 85'F
.. 1(S
- m. . #.WEUTW W n55E sa.
lS!( y '. 1 Q.3 Q .'?w-] _i w?
- n y't , - g;Q;.f" '
a, 2A.. .L p.. - "$ 4 bl2: -:y@. i
. :. r
- c. c ms-r
-~ m m. . . . . . .- . p4 9 J % h, . 1j; b > %.- Lpg?D., c
$$W ' lW V;Ql, ddd&bf 1
,.n .- , . . =,
)
m
').;;^: b2:M;gjgsghg,Q l
~ ;y;2 p q% my; W;1 ,
M e W . T;;:t :;.- sg A s;0 %.,:.:.,ulu.L i.+; i t L;b ?]n. $ lE A h N dCl' Specimen AL14 200*F mg (,~,~
.;- yyga y 1s m s. e. 4 ex .
*2's ' Q : - Vu
- .., , op: q - e c o. o31
^s "X.
. - ' q ; *(s y ' .
}. < gp r s q qq :_ y
'"x .& 3f:3, , ::%
- x- ' M, n .. 4. . c 4.u; c +
, e
.p _< s
~ ~~.. , ... ,. - .-- - ,
, ~w . .
Specimen AL15 550*F 4 Figure 5-12 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) i i 5-29 l
9_ i i 1 J 4 9 i r s i m , ,4 +0 u. 4'eyW t p,,3 N -
%ay, .
- 3 . q
- , . .: s 4 sst >
gg v n;a gn;n ;yp%T 4>w - e.g, .. p;ek ,
. :3 : a, ,, ,
- _:nq7c 30%; s l f & =l. q' 2
WdkN(}dNh;g}2,1;rkfM;AIMM$, MD YON $Q f.i
- W w~ .,;4gg ;%, ,2;;;:? ' . WW jy 5 ,,. ~ +' 1%x.,
n. x +i<
< : ,g , _', ;, & . c. ,
p ,_.
- EN' 41/t M M k S M b i%ddU dl. k N A + a s O 2. d al'. M ^4 1 Specimen AT13 100*F 1
i W 1 (:; M3 vt .:: ; ;p*M%fi: 2 -n'aga [n l 'B0 6:l ; 1i W3N4)g.c 0:fi'. .. ' D
'p" i, afisg hy:9 [
] - 3 h: lQn d;591 W W, b /! I 'IM61/MN;;$ DON @SIMIM 1 i - 1 . jg
- n. 7 r l ?
. g ~.u..pg
- e- ,
y.
- *1 r*, '
c A:a o,.
.+. A D s L M,4Ty p ey~a'
-A s 9 /shme un..w:4~.u. -
i Specimen AT14 225'F ) A j l i ,
;,wg.cg'* 4ry'.pr,,
s-+-c%_..- j 3 m 4 r - . r a ,. ,r y -
't e,w M t .
T p,; .g g 1' i
>'h
.4) l 8 ^
wgg. . n4 "
.yy g i
t ( Qvjg.l's
- 1
.(
-, . y.k*$q t'" ' '
n, .: . ( f
,1 \
i a
.~
#w &
s
.a, u.% A ;_.
9;~. ~ s Specimen AT15 550'F ' A
- i t l j -
4 i e 1 g Figure 5-13 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) h i } 5-30 ! i l
-~ . . . ~ . - . . . - - -
- ~- .
4 A t 4 I i e i i 1 4
' v '; y~ '. : e .- ;; s.
- y2
, r. ,- 3 . . y, w,c.r,r? A :: r:.3 '
9 - 5
.%i: + .f r f , ., g. ,,,.,.. ,/.1,pp ' .,
s wu ,. .e r
- ww.
- = i g . .
- -
'Y % l ;
) , .. Q .. , s v., 9.'re :.r?.;. y ; .<-
*4 ..
5 c.s..
- ga y W. ( ,
- 2
- . e 4, .. .
4 .
,s w f b s' 1 +
4 w c,.m m ;, a., n. . , t t I
-y"Y' **l* QW% jk
~ .x
, m -p; , ,
9994~.] w-[ ,. -4 C..r.aGk.2cG. a /al;;.d;.5 : gia i,,3, ' jgg Q},gf . i Specimen AW13 175'F i. 1 i f 1 i .. -
- y . p. .
f {@. . , . - .Qh ___j j ,
.y . g((. , ~. ., m .jjy t' ,
,y > a:6m ,
g; y 1 gf' * ' > , u-. ${AW M Q A ! $ 4, I 1 C,T . ' $. ' (, '
%v
' . 4 1 x 1@ 1, a 4 e _mzg c 4 a o_. .--..e~- s ,. . , 1 4 j -- i 4.1 - 5.La-a Specimen AW14 25* F ) i 4 4 4 3 4
- yw ,, _..,.s_,. .. _ _ _ _ _ _ _
l A - J : ..
.:, Q ,
a yh
;.cw;g
../,. L <A w-J '
;/c ;. %J -
MTG.c 7'
,;.- L; r5 ~^; QY. :.~ y'9>..,%.
v
,s
' y.
I .
-u. ,..a 3 w ..
- ...,g.
, s s.
. ..e.
t n n, s . .o.. ga. wgf. ngp p ; ,
. A a
3 . , , .,c,- v $ ' ,:b.L ' 'if b:j i7. ,y.f ,. , 4 Q 7. . t ds 1-a3 g-g. ~/ ;3 ~ J i e 4 s (h .i C, 4 g g'y . y .. , . , ,
.h,c_ ;n .
\
i j ,
, +
- - . ..y e --' 3:npy, y j
, ,w & - ~ .s
.na . . ,
. c ,,m.. .
! Specimen AW15 550'F 1 i
i f i - 1 i i i Figure 5-14 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel 1 Surveillance Weld Metal i i 1 4 1 5-31 1 _, - _ _ _
STo.ESS-STRAIN CURVE VOGTLE UNIT 1 "Y" CAPSULE 100.00 , 90.00- y' ^ s,
~
80.00- / \ g 70.00 Y - 60.00-CD_
@ 50.00-{
40.004 . m l 30.001 AL13 20.00-85 F 10.00j I 0.00 0.00 0.10 0.20 STRAIN, IN/IN STHESS-Si~ RAIN CUHVE VOGTLE UNIT 1 "Y" CAPSULE 100.00 90.00- - 80.00-g 70.00-M 60.00-cn_
@ 50.00-b 40.00-U .
30.00-20.00- AL 14 10.00- 200 F 0.00 , , , . 0.00 0.10 0.20 l STRAIN, IN/IN
-l l
l Figure 5-15 Engineering Stress-Strain Curves for Intermediate Shell Plate B8805-3 Tensile Specimens AL13 and ALl4 (longitudinal Orientation) 5-32
STRESS-STRAIN CURVE VOGTLE UNIT 1 "Y" CAPSULE 100.00 90.00-80.00-70.00- _o
- 60.00-l 6 l @ 50.00-c:
$ 40.00-30.00-AL15 20.00- ~
550 F 10.00-l 0.00 . . !- 0.00 0.10 0.20 STRAIN, IN/IN 1 l l l l l l i ?- , i Figure 5-16 Engineering Stress-Strain Curve for Intermediate Shell Plate B8805-3 I Tensile Specimen AL15 (Longitudinal Orientation) 1 l 5-33 t
STRESS-STRAIN CUHVE' VOGTLE UNIT 1 "Y" CAPSULE 100.00
. _ . . ~
90.00- N, 80.00- [,[ - , g 70.00-M 1 60.00-cD_ ,
@ 50.00-I i H 40.00-U) 30.00-AT 13 20.00-10.00-0.0 '
0.00 0.10 0.20 - STRAIN, IN/IN STRESS-STRAIN CURVE VOGTLE UNIT 1 "Y" CAPSULE 100.00 -l
.90.00-80.00-70.00-55 x 60.00-
US
$ 50.00- .
a:
$ 40.00- l 30.00-20.00-AT 14 10.00-225 F 0.00 . . , ,
0.00 0.10 0.20 STRAIN, IN/IN Figure 5-17 Engineenng' Stress-Strain Curves for Intermediate Shell Plate B8805-3 Tensile Specimens AT13 and AT14 (Transverse Orientation) 5-34
1 STRESS-STRAIN CURVE VOGTLE UNIT 1 "Y" CAPSULE 100.00 90.00-80.00-70.00-m
- 60.00-vi G 50.00-a-
$ 40.00-30.00-20.00- AT 15 10.00 550 F I
o.co , , , , , . , , 0.00 0.04 0.08 0.12 0.16 0.20 STRAIN, IN/IN Figure 5-18 Engineering Stress-Strain Curve for Intermediate Shell Plate B8805-3 Tensile Specimen ATIS (Transverse Orientation) 5-35 1
l STRESS-STRAIN CURVE ! VOGTLE UNIT 1 'Y" CAPSULE. ; 100.00 90.00- ... g 8o o~
'70.00-
,f ,. N N 60.00-
. l I
@ 50.00-CC :
H 40.00-to 30.00- . 20.00- AW 13 10.00- 175 F 1 0.00 . . , . 0.00 0.10- 0.20 STRAIN, . IN/IN STRESS-STRAIN CURVE . VOGTLE UNIT 1 "Y" CAPSULE '
- i 100.00 90.00- _
80.00- i 70.00-e j
- 60.00-O,
@ 50.00-c y 40.00- ,
l 30.00-l AW 14 20.00-l 25 F i 10.00- ! 0.00 . . . . . 0.30 I 0.00 0.10 0.20 l STRAIN, IN/IN i 1 I Figure 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specunens AW13 and AW14 j ! 5-36 I
STRESS-STRAIN CURVE VOGTLE UNIT 1 "Y" CAPSULE 100.00 90.00-80.00-70.00- _u)
- 60.00-vi
@ 50.00-e y 40.00-30.00-20.00- AW 15 l' 10.00- 550 F 0.00 . . .
1 0.00 0.10 0.20 STRAIN, IN/IN 1 i 1 l l l j Figure 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimens AW15 I ! 5-37
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the pressure vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is generally derived solely from analysis. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for development of 14 mage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between smveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall. Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reponing displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing
, Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials."
6-1
This section provides the results of the neutron dosimetry evaluations performed in conjunction with 1 1 the analysis of test specimens contained in surveillance cipsule Y, withdrawn at the end of the fourth fuel cycle. Also included is an updated evaluation of the dosimetry contained in capsule U, withdrawn at the conclusion of cycle one. This update is based on current state-of-the-art methodology and ^l nuclear data; and, together with the capsule Y results, provides a consistent up to date data base for use in evaluating the material properties of the Vogtle Unit I reactor vessel. ! In each of the dosimetry evaluadons, fast neutron exposure parameters in terms of neutron fluence (E i i > 1.0 MeV), neutron fluence (E > 0.1 MeV), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel wall. Also, uncertainties associated with the derived exposure parameters at the surveillance capsules and , with the projected exposure of the pressure vessel are provided. 6.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 58.5 ,61.0 ,121.5 , 238.5*,241.0 , and 301.5' relative to the core cardinal axis as shown in Figure 4-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1. The stainless steel specimen containers are 1.182 by 1 inch and approximately 56 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 feet of the 12 foot high reactor core. From a neutronic standpoint, the surveillance capsules and associated support structures are significant. The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model. In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the 6-2
1 d conventional forward mode, was used primarily to obtain relative neutron energy distributions l throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {$(E > 1.0 MeV),$(E > 0.1 MeV), and dpa/sec} through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the
~
surveillance capsules as well as for the determmation of exposure parameter ratios; i.e., [dpa/sec]/[$(E
> 1.0 MeV)), within the pressure vessel geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T,1/2T, and 3/4T locations.
- The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux,
$(E > 1.0 MeV), at smveillance capsule positions and at several azimuthallocations on the pressure vessel inner radius to neutron source distributions within the reactor core. The source importance j functions generated from these adjoint analyses provided the basis for all absolute exposure calculations and comparison with measurement. These imponance functions, when combined with fuel cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation. They also established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core but also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel i
- assemblies increased.
The absolute cycle specific data from the adjoint evaluations together with the relative neutron energy i spectra and radial distribution information from the reference forward calculation provided the means l to: 1 1- Evaluate neutron dosimetry obtained from surveillance capsules.
- 2- Extrapolate dosimetry results to key locations at the inner radius and through the thickness of f the pressure vessel wall.
3- Enable a direct comparison of analytical prediction with measurement. 4- Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves. J 6-3 l l I
The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R,0 geometry using the DOT two-dimensional discrete ordinates codeUU and the SAILOR cross-section library!"). The SAILOR library is a 47 energy group ENDF/B-IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was treated
~
with a P 3expansion of the scattering cross-sections and the angular discretization was modeled with an S, order of angular quadrature. The core power distribution utilized in the reference forward transport calculation was derived from statistical studies of long-term operation of Westinghouse 3-loop plants. Inherent in the development . of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, the neutron source was increased by a 2o margin derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power. Since it is unlikely that any single reactor would exhibit power levels on the core periphery at the nominal + 2a value for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results. All adjoint calculations were also carried out using an S, order of angular quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source locations were chosen at several ~ azimuthal locations along the pressure vessel inner radius as well as at the geometric center of each surveillance capsule. Again, these calculations were mn in R.0 geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case $(E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as: R(r,0) = f [ f 1(r,0,E) S(r,0,E) r dr de dE r eE where: R(r,0) = 4(E > 1.0 MeV) at radius r and azimuthal angle 0. I(r,0,E)= Adjoint source importance function at radius r, azimuthal angle 0, and neutron source energy E. S(r,0,E)= Neutron source strength at core location r,0 and energy E. Although the adjoint importance functions used in this analysis were based on a response function defined by the threshold neutron flux 4(E > 1.0 MeV), prior calculations"U have shown that, while the 6-4
implementation of low leakage loading patterns significantly impacts both the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order, j- Thus, for a given location the ratio of [dpa/sec]/[$(E > 1.0 MeV)] is insensitive to changing core source distributions. In the application of these adjoint imponance function.; to the Vogtle Unit 1 reactor, therefore, the iron atom displacement rates (dpa/sec) and the neutron flux $(E > 0.1 MeV) were computed on a cycle specific basis by using [dpa/sec]/[$(E > 1.0 MeV)] and [$(E > 0.1 MeV)]/[$(E > 1.0 MeV)] ratios from the forward analysis in conjunction with the cycle specific $(E > 1.0 MeV) solutions from the individual adjoint evaluations. The reactor core power distributions used in the plant specific adjoint calculations were taken from the fuel cycle design repons for the first four operating cycles of Vogtle Unit 1 D'*"*** . Selected results from the neutron transport analyses are provided in Tables 6-1 through 6 5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the pressure vessel wall. In Table 6-1, the calculated exposure parameters [$(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec] are l-given at the geometric center of the two surveillance capsule positions for both the reference and the plant specific core power distributions. The plant specific data, based on the adjoint transpon analysis, are meant to establish the absolute comparison of measurement with analysis. The reference data derived from the forward calculation are provided as a conservative exposure evaluation against which plant specific fluence calculations can be compared. Similar data are given in Table 6-2 for the pressure vessel inner radius. Again, the three peninent exposure parameters are listed for the reference ; i : ! and the cycle one through four plant specific power distributions. It is imponant to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the ! l maximum predicted exposure levels of the vessel wall itself. i l I Radial gradient information applicable to $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec is given in l Tables 6-3,6-4, and 6-5, respectively. The data, obtained from the reference forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Tables 6-3 l through 6-5, 6-5 l t
For example, the neutron flux 4(E > 1.0 MeV) at the 1/4T depth in the pressure vessel wall along the 45' azimuth is given by: I
-I
$gf45') = $(220.27, 45 ) F(225,75, 45o) ,
where: 4m(45') = Projected neutron flux at the 1/4T position on the 45* azimuth. 4(220.27,45') = Projected or calculated neutron flux at the vessel inner radius on the 45 azimuth. F(225.75,45') = Ratio of the nutron flux at the 1/4T position to the flux at the vessel inner radius for the 45 azimuth. This data is obtained from Table 6-3. Similar expressions apply for exposure parameters expressed in terms of 4(E > 0.1 MeV) and dpa/sec I where the attenuation function F is obtained from Tables 6-4 and 6-5, respectively. i l 6.3 Neutron Dosimetry The passive neutroa sensors induded in the Vogtle Unit I surveillance program are listed in Table 6-6. l Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were - used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the
~
subsequent determination of the various exposure parameters of interest [4(E > 1.0 MeV),4(E > 0.1 MeV), dpa/sec]. The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes I drilled in spacers at several axiallevels within the capsules. The cadmium shielded uranium and neptunium fission monitors were accommodated within the dosimeter block located near the center of the capsule. The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: The measured specific activity of each monitor. 6-6
l
- The physical characteristics of each monitor.
The operating history of the reactor. 1- - The energy response of each monitor.
- The neutron energy spectrum at the monitorlocation.
l The specific activity of each of the neutron monitors was determined using established ASTM I procedurest24 *"* M. Following sample preparation and weighing, the activity of each monitor was i l determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation ! history of the Vogtle Unit I reactor during cycles one through four was supplied by NUREG-0020, 1 ! " Licensed Operating Reactors Status Summary Report." for the applicable period. The irradiation history applicable to capsules Y and U is given in Table 6-7. Note that the Vogtle Unit I uprating / T-hot reduction has increased the full power design reactor power rating from 3411 to 3565 MWt. The analysis perfonned for capsule Y used the new power rating for all calculations. l Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full power operation were determined from the following equation: A
~
R, P NFYE p l C j[1-e "/] [e-"d] o ret where: R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P, 'r e/ nucleus). A = Measured specife activity (dps/gm). No = Number of target element atoms per gram of sensor. F = Weight fraction of the target isotope in the sensor material. Y = Number of product atoms produced per reaction. P3 = Average core power level during irradiation period j (MW). Pg = Maximum or reference power level of the reactor (MW). C3 = Calculated ratio of $(E > 1.0 MeV) during irradiation period j to the time weighted average &(E > 1.0 MeV) over the entire irradiation period. A = Decay constant of the product isotope (1/sec). t, = Length of irradiation period j (sec). t, = Decay time following irradiation period j (sec). 6-7
and the summation is carried out over the total number of monthly intervals comprising the irradiation l period. i In the equation describing the reaction rate calculation, the ratio [P)/[P,e] accounts for month by month variation of reactor core power level within any given fuel cycle as well as over multiple fuel - l cycles. The ratio C,3 which can be calculated for each fuel cycle using the adjoint transport technology 1 1 discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux i level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single cycle irradiation C3 is normally taken to be 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel management, the additional C3 term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for rnany cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved l from one capsule location to another. For the inadiation history of capsules Y, the flux level term in the reaction rate calculations was developed from the plant specific analysis provided in Table 6-1. Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Table 6-8. - Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least quares adjustment code"H. The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data. The " measured" exposure parameters along with the associated uncertainties were l l then obtained kom the adjusted spectrum. I i l In the FERRET evaluations, a log-normalleast squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncenainties and correlations. In general, the measured values f are linearly related to the flux $ by some response matrix A: f = E A lg" Q 8 - where i indexes the measured values belonging to a single data set s, g designates the energy group, and a delineates spectra that may be simultaneously adjusted. For example, - 6-8 i
h"ECighg 8 relates a set of measured reaction rates Ri to a single spectmm $, by the multigroup reaction cross- .' section o,,. The log-normal approach automatically accounts for the physical constraint of positive 1 l fluxes, even with large assigned uncecainties. l ( In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were l approximated in a multi-group format consisting of 53 energy groups. The trialinput spectrum was converted to the FERRET 53 group structure using the SAND-II code *1 This procedure was carried l out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET. l The sensor set reaction cross-sections, obtained from the ENDF/B-V dosimetry file, were also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting function in the cross-section collapsing procedure. Reaction cross-section uncenainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constmeted from the information contained on the EhTF/B-V data files. These matrices included energy group to energy group uncertainty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were not included. The omission of this additional uncertainty information does not significantly impact the results of the adjustment. Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the seraar set locations, the neutron spectrum input to the FERRET evaluation was taken from the center of the surveillance capsule modeled in the reference forward transport calculation. While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the ENDF/B-V data files, the covariance matrix for the input trial spectrum was constructed from the following relation: 6-9
i where R, specifies an overall fractional normalization uncertainty (i.e., complete coiTelation) for the set of values. The fractional uncertainties R, specify additional random uncertainties for group g that are correlated with a correlation matrix given by: . Py, = [1-0] 5y, + 0 e -" , H = @-8Y 2 2y The first term in the correlation matrix equation specifies purely random uncertainties, while the l second term describes short range correlations over a group range y (0 specifies the strength of the latter term). The value of 6 is I when g = g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerke/*. Maerker's results are closely duplicated when y = 6. The uncertainties associated with the measured reaction rates included both statistical (counting) and systematic components. The systematic component of the overall uncertainty accounts for counte! - efficiency, counter calibrations, irradiation histoi,/ corrections, and corrections for competing reactions in the individual sensors. Results of the FERRET evaluations of the capsule Y and U dosimetry are given in Table 6-12. The data summarized in this table include fast neutron exposure evaluations in terms of $(E > 1.0 MeV),
*(E > 0.1 MeV), and dpa. In general good results were achieved in the fits of the adjusted spectra to the individual measured reaction rates. The adjusted spectra from the least squares evaluations are given in Tables 6-11 in the FERRET 53 energy group structure. The results for capsule Y are consistent with results obtained from similar evaluations of dosimetry from other Westinghouse reactors.
l 6.4 Proiections of Pressure Vessel Exposure Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-
- 13. Along with the current (4.64 EFPY) exposure, projections are also provided for exposure periods of 16 EFPY and 32 EFPY. In computing these vessel exposures, the calculated values from Table 6-2 6-10 l
l
were scaled by the average measurement / calculation ratios (M/C) observed from the evaluations of dosimetry from capsules Y and U for each fast neutron exposure parameter. This procedure resulted in bias factors of 1.08,1.04, and 1.02 being applied to the calculated values of $(E > 1.0 MeV), &(E
> 0.1 MeV), and dpa, respectively. Projections for future operation were based on the assumption that the average exposure rates characteristic of the cycle one through four irradiation would continue to be applicable throughout plant life.
The overall uncertainty associated with the best estimate exposure projections at the pressure vessel wall depends on the individual uncenainties in the measurement process, the uncenainty in the dosimetry location, and on the uncertainty in the extrapolation of results from the measurement points : to the point of interest in the vessel wall. For Vogtle Unit 1, the uncenainty in each individual capsule derived fluence is estimated to consist of a 6% random component and a 5% systematic component, and the extrapolation uncertainty is estimated to be 5%. A statistical combination of these uncenainties for the two capsules produces an overall uncertainty estimate in the exposure of the pressure vessel wall in the beltline region of 8% (lo) for @(E > 1.0 MeV). In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Vogtle Unit I reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were also employed. Data based on both a $(E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14. In order to define RTxm, from existing correlation curves vs fluence (E>1.0 MeV), dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations: h
+(V4D = 4(0D a(OD @"(Y4D
""4 4
4(y4D = 4(oD @a(W4D a(OD Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Table 6-15
. updated lead factors are listed for each of the Vogtle Unit I surveillance capsules.12ad factor data 4 based on the accumulated fluence through cycle four are provided for each remaining capsule.
6-11
FIGURE 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule
. 1 i
l . l . l l (TYPICAL) . j C'
- 88.5' -s1.08
- s\
h // 2m&l: 0 4 6-12 l
FIGURE 6-2 Axial Distribution of Neutron Fluence (E > 1.0 hEV) Along the 45 Degree Azimuth ! 1.0E+20 I I l l 1.0E+19 - -- - A E
,- .'s E f ^ s.
i
~ 1.0E+18
/ \
G O C D
.2 u_
c 1.0E+17 2 3@
. Z 1.0E+16
-4.64 EFPY -- 15 EFPY 32 EFPY
~
1.0E+15 "" ' "' 0 1 2 3 4 5 67 8 9 101112 ; Distance From Core Bottom (ft) 6-13
i i TABLE 6-1 1 i Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center l . ~ ) 2 CALCULATED FLUX & (E > 1.0 MeV [n/cm -sec]) AT THE SURVEILLANCE CAPSULES . [ CAPSULE LOCATION ! 29.0 31.5' l a. j CYCLE 1 8.946E+10 9.537E+10 . i CYCLE 2 7.634E+10 8.131E+10 .; i- CYCLE 3 7.806E+10 8.570E+10 t [ CYCLE 4 6.359E+10 6.857E+10 ; } CRSD Data 1.130E+11 1.210E+11 ? 2 q CALCULATED FLUX $ (E > 0.1 MeV (n/cm -sec]) AT THE SURVEILLANCE CAPSULES i 29.0' ' 31.5' > l ) CYCLE 1 3.869E+11 '4.089E+11 I l CYCLE 2 3.302E+11 3.486E+11 i 1 CYCLE 3 3.376E+11 3.674E+11 l CYCLE 4 2.750E+11 2.940E+11 CRSD Data 4.887E+11 5.187E+11 - CALCULATED Iron Displacement Rate [dpa/sec] AT THE SURVEILLANCE CAPSULES ~ 29.0 31.5 CYCLE 1 1.753E-10 1.860E-10 CYCLE 2 1.4%E-10 1.586E-10 : CYCLE 3 1.530E-10 1.671E-10 CYCLE 4 1.246E-10 1.337E-10 CRSD Data - 2.215E-10 2.360E-10
~
l 6-14 < l 1
-l l
. . . . . - - = . ~ -- - - . _ . .- ---.-.- _. -. .. .. . . . _ _
i TABM 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates
- ' at the Pressure Vessel Clad / Base Metal Interface t
6(E > 1.0MeV) In/cm 2-secl I L 0* 15
- 25
- 35' 45'
.. CYCG 1 1.407E+10 - 2.077E+10 2.378E+10 1.933E+10 2.209E+10 CYCE 2 1.094E+10 1.708E+10 2.011E+10 1.616E+10 1.762E+10 CYCW 3 9.943E+09 1.550E+10 1.964E+10 1.719E+10 1.903E+10 ,
*~
CYCW 4 1.110E+10 1.508E+10 1.693E+10 1.406E+10 1.614E+10 - CRSD Data 1.780E+10 2.660E+10 3.010E+10 2.450E+10 2.810E+10 2 4(E > 0.1MeV) In/cm -secl - > 0* 15
- 25
- 35' 45' '
! CYCG 1 2.925E+10 4.373E+10 6.494E+10 -5.491E+10 5.534E+10 l CYCG 2 2.274E+10 3.5%E+10 5.492E+10 4.591E+10 4.414E+10 l CYCE 3 2.%7E+10 3.263E+10 5.363E+10 4.883E+10 4.768E+10 CYCM 4 2.307E+10 '3.175E+10 4.623i3+10 3.994E+10 - 4.044E+10 CRSD Data 3.700E+10 5.600E+10 8.220E+10 6.960E+10 7.040E+10 Iron Atom Displacement Rate idea /secl . l- 0 15 ' 25 * ' 35' 45 CYCW l 2.190E-11 3.217E-11 3.982E-11 ' 3.274E-11 3.522E-11 i CYCB 2 1.702E-11 2.645E-11 3.367E-11 2.737E-11 2.809E-11 CYCE 3 1.547E-11 2.401E 3.289E-11 2.912E-11 3.034E-11
-CYCE 4 .1.727E-11 2.336E-11 2.835E ' 2.382E-11 2.573E CRSD Data 2.770E-11 4.120E-11 5.040E-11 4.150E-11 4.480E-11 l
l l l 6-15
l TABLE 6-3 Relative Radial Distribution of $(E > 1.0 MeV) within the Pressure Vessel Wall Radius (cm) 0. 0 15.0 25.0 35.0 45.0 220.27m 1.00 1.00 1.00 1.00 1.00 220.64 0.976 0.979 0.980 0.977 0.979 221.66 0.888 0.891 0.893 0.891 0.889 - 222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 0.554 0.552 0.543 226.95 0.462 0.460 0.465 0.463 0.452 228.28 0.386 0.384 0.388 0.386 0.375 229.60 0.321 0.319 0.324 0.321 0.311 230.92 0.267 0.263 0.275 0.267 0.257 232.25 0.221 0.219 0.225 0.221 0.211 i 233.57 0.183 0.181 0.185 0.183 0.174 234.89 0.151 0.149 0.153 0.151 0.142 236.22 0.124 0.122 0.126 0.124 0.116 237.54 0.102 0.100 0.104 0.102 0.0M5 - 238.86 0.0828 0.0817 0.0846 .0835 0.0762 240.19 0.0671 0.0660 0.0689 .0679 0.0608 241.51 0.0538 0.0522 0.0550 0.0545 0.0471 242.17* 0.05 % 0.0488 0.0518 0.0521 0.0438 1 NOTES: 1) Base Metal Inner Radius i
- 2) Base Metal Outer Radius i
6-16
l TABLE 6-4 Relative Radial Distribution of $(E > 0.1 MeV) within the Pressure Vessel Wall Radius (cm) 0. 0* 15.0* 25.0* 35.0* 45.0 l 220.27m 1.00 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 1.00
, 221.66 1.00 1.00 1.00 0.999 0.995
-222.99 . 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901-
! 225.63 0.874 0.865 0.874 . 0.850 0.842 226.95 0.818 0.808 0.818 0.792 0.782 228.28 0.761 0.750 0.716 0.734 0.721 229.60 0.705 0.693 0.704 0.677 0.662 230.92 0.649 0.637 0.649 0.621 0.605 232.25 0.594 0.582 0.594 0.567 0.549 233.57 0.540 0.529 0.542 0.515 0.495 . 234.89 0.487 0.478 0.490 0.465 0.443 236.22 0.436 0.428 0.440 0.416 0.392
- 237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 - 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201 242.175 0.233 0.226 .237 0.223 0.188-NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius n
I 6-17
i TABLE 6-5 Relative Radial Distribution of dpa/see within the Pressure Vessel Wall Radius l (cm) 0. 0' 15.0 25.0 35.0 45.0* 220.27W 1.00 1.00 1.00 1.00 1.00 220.64 0.984 0.981 0.984 0.983 0.984 l 221.66 0.912 0.909 0.917 0.921 0.915 - 222.99 0.815 0.812 0.826 0.833 0.821
~
224.31 0.722 0.719 0.737 0.747 0.730 225.63 0.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572 228.28 0.497 0.493 0.519 0.533 0.506 i 229.60 0.439 0.435 0.462 0.475 0.447 230.92 0.387 0.383 0.410 0.423 0.3 M 232.25 0.341 0.338 0.364 0.376 0.347 233.57 0.300 0.297 0.322 0.334 0.305 234.89 0.263 0.261 0.285 0.295 0.266 236.22 0.230 0.228 0.250 0.260 0.231 237.54 0.199 0.198 0.218 0.227 0.199 - 238.86 0.171 0.170 0.189 0.1% 0.169 240.19 0.145 0.144 0.161 0.167 0.140 241.51 0.121 0.119 0.135 0.139 0.113 242.175 0.116 0.113 0.128 0.134 0.106 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius l
1 a 6-18
I TABLE 6-6 Nuclear Parameters Used In the Evaluation of Neutron Sensors Reaction Target Fission Monitor of Weight Response Product Yield Material Interest Fraction Range Half-Life (%) Copper Cu"(n,a)Co" 0.6917 E > 4.7 MeV 5.271 yrs Iron Fe"(n.p)Mn" 0.0580 E > 1.0 MeV 312.5 days
. Nickel NiS8(n.p)Co 58 0.6827 E > 1.0 MeV 70.78 days 2
Uranium-238* U "(n f)Cs"' 1.0 E > 0.4 MeV 30.12 yrs 6.00 Neptunium-237* Np237(n,f)Cs"7 1.0 E > 0.08 MeV 30.12 yrs 6.27 Cobalt-Aluminum
- Co"(n,y)Co* 0.0015 0.4ev>E> 0.015 MeV 5.271 yrs Cobalt-Aluminum Co"(n,y)Co" 0.0015 E > 0.015 MeV 5.271 yrs
- Denotes that monitor is cadmium shielded.
1 ea 6-19
TABLE 6-7 Monthly Thermal Generation During the First Four Fuel Cycles of the Vogtle Unit 1 Reactor Thermal Generation Thermal Generation Year Month (MW-hr) Year Month (MW-hr) 1987 3 68,766 1991 1 2,534,837 4 797,491 2 2,260,779 5 1,044,332 3 2,495,386 6 759,746 4 2,449,552 7 1,835,718 5 2,533,685 8 2,509,822 6 2,449,889 9 2,452,829 7 2,534,501 10 707,673 8 2,483,204 - 11 1,927,388 9 969,976 12 2,467,702 10 0 1988 1 1,365,280 1 215,953 2 1,387,377 2 2,466,013 3 2,456,340 1992 1 2,534,684 4 1,907,244 2 2,371,364 5 2,531,355 3 2,528,590 6 2,444,967 4 2,239,948 7 2,220,349 5 1,866,712 8 2,415,264 6 2,452,840 9 2,370,737 7 2,534,681 10 483,956 8 2,535,008 11 52,233 9 2,188,889 12 2,135,007 10 2,538,900 1989 1 1,771,903 11 2,454,211 2 1,905,573 12 2,536,190 3 2,533,004 1993 1 2,536,730 4 2,380,073 2 2,273,143 5 2,264,902 3 849,752 , 6 2,452,382 7 2,443,387 Cumulative 1.449E+08 8 2,286,024 9 2,450,229 10 2,142,954 11 2,391,716 12 2,535,607 1990 1 2,374,089 2 1,811,171 3 0 4 591,136 5 2,311,713 6 2,299,026 7 2,196,834 8 2,512,580 9 2,452,206 10 2,534,258 11 2,428,733 12 1,692,955 w 6-20
( l TABLE 6-8 Measured Sensor Activities and Reaction Rates
. Surveillance Capsule Y Saturated Activities and Derived Fast Neutron Flux MEASURED SATURATED RFACTION MONITOR AND ACTIVITY ACTIVITY RATE AXIAL LOCATION (dis /sec-sn) (dis /sec mn) (rps/ nucleus)
Cu-63 (n.a) Co-60 l 93-3522 TOP 1.380E+05 3.684E+05 93-3527 MID 1.210E+05 3.230E+05 l 93-3532 BOT 1.230E+05 3.284E+05 , l AVERAGES 1.273E+05 3.400E+05 5.I86E-17 l l Fe-54 (n.n) Mn-54 ( 93-3524 TOP 1.630E+06 3.222E+06 l 93 3529 MID 1.470E+06 2.906E+06 ( _93-3534 BOT 1.480E+06 2.925E+06 l AVERAGES 1.527E+06 3.018E+06 4.825E-15 Ni 58 (n.9) Co-58 93-3523 TOP 8.430E+06 5.031E+07 93-3528 MID 7.750E+06 4.625E+07 93-3533 BOT 7.630E46 4.553E+07 AVERAGES 7.937E+06 4.736E+07 6.763E-15 Co-59 (n y) Co-60 93-3520 TOP 2.340E+07 6.247E+07 93 3525 MID 2.350E+07 6.274E+07 93 3530 BOT 2.340E+07 6.247E+07 AVERAGES 2.343E407 6.256E+07 4.082E-12 Co-59 (n.v) Co-60 93-3521 TOP 1.200E+07 3.204E+07 93-3526 MID 1.290E+07 3.444E+07 93-3531 BOT 1.290E+07 3.444E+07 AVERAGES 1.260E+07 3.364E+07 2.195E-12 U-238 (n.O Cs-137 93-3518 MID 5.070E+05 5,473E+06 3.607E-14 1 No-237 (n.O Cs-137 93 3519 MID 3.380E+06 3.649E407 2.291E-13 6-21
, _ _ - _ i_ . _,__..- _. _ - , . .
4 l I TABLE 6-9 , Summary of Neutron Dosimetry Results ! Surveillance Capsules Y and U - i Calculation of Measured Fluence for Capsule Y Flux Time Fluence Uncertainty 1 1 Meas Fluence < 0.414 ev = (Meas Flux <0.414) * (EFPS) 7.696E+10 1.464E+08 1.127E+19 t22% l Meas Fluence > 0.1 Mev = (Meas Flux > .1) * (EFPS) 3.415E+11 1.464E+08 5.000E+19 115 % Mens Fluence > 1.0 Mev = (Meas Flux > 1) * (EFPS) 8.484E+10 1.464E+08 1.242E+19 iS% , dpa 1.542E 10 1.464E+08 2.258E-02 *11% ; Calculation of Measured Fluence for Capsule U Flux Time . Fluence Uncertainty Meas Fluence < 0.414 ev = (Meas Flux < 0.414) * (EFPS) 1.035E+11 ' 3.449E+07 3.5698+18 121 % Meas Fluence > 0.1 Mev = (Meas Flux > .1) * (EFPS) 4.296E+11 3.449E+07 1.481E+19 *15% Meas Fluence > 1.0 Mev = (Meas Flux > 1) * (EPPS) 9.967E+10 3.449E+07 3.437E+18 18 %
.1 dpa 1.882E-10 3.449E+07 6.490E-03 ti1%
l _) l l I ,
+
l i l l e e 6-22 l
I l i l: TABLE 6-10 i Comparison of Measured and FERRET Calculated Reaction Rates at the Stuveillance Capsule Center l- Stuveillance Capsule Y l ADJUSTED REACTION hEASURED CALCULATION C/M 'l Cu-63 (n a) Co-60 5.19E-17 5.26E-17 1.01 l Fe-54 (n.p) Mn-54 4.83E-15 4.87E-15 1.01 : Ni-58 (n.p) Co-58 6.76E-15 6.7IE-15 0.99 U-238 (n.f) Cs-137 (Cd) 3.01E-14 2.71E 4 0.90 NP237(N,F)CS137 2.29E-13 2.49E-13 1.09-Co-59 (n;y) Co-60 4.08E-12 4.05E-12 0.99 Co-59 (n,y) Co-60 (Cd) 2.20E-12 2.21E-12 1.0 l l l l - i e'- 6-23
TABLE 6-11 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule Y ; 1
- ENERGY ADIUSTED FLUX ENERGY ADJUSTED FLUX 2
! GROUP (MeV) (n/cm'-sec) GROUP (MeV) (o/cm -sec) . 1 1.733E+01 7.406E+06 27 1.503 S 02 1.289E+10 ) i 2 1.492E+01 1.678E+07 28 9.119E-03 1.6ME+10 3 1.350E+01 6.479FA07 29 5.531E-03 1.910E+10 4 1.162E+01 1.444E+08 30 3.355E-03 6.083E409 5 1.000E+01 3.189E+08 31 2.839E-03 5.914E409 6 8.607E+00 5.427E+08 32 2.404E-03 5.821E+09 l 7 7.408E+00 1.241E+09 33 2.035E-03 1.721E+10 8 6.065E+00 1.757E+09 34 1.234E-03 1.660E+10 l 9 4.966E+00 3.635E+09 35 7.485E-M 1.481E+10 10 3.679E+00 4.731E+09 36 4.540E-N 1.286E+10 11 2.865E+00 9.785E409 37 2.754E-M 1.491E+10 12 2.231E+00 1.322E+10 38 1.670E-04 1.653E+10 13 1.738E+00 1.799E+10 39 1.013E-04 1.638E+10 14 1.353E400 1.941E+10 40 6.144E-05 1.617E+10 15 1.108E+00 3.506E+10 41 3.727E-05 1.592E+10 16 8.208E-01 3.806E+10 42 2.260E-05 1.546E+10 17 6.393E-01 4.026E+10 43 1.371E-05 1.492E+10 18 4.979E-01 2.742E+10 44 8.315E-06 1.422E+10 19 3.877E-01 3.932E+10 45 5.043E-06 1.328E+10 20 3.020E-01 4.218E+10 46 3.059E-06 1.237E+10 21 1.832E-01 3.987E+10 47 1.855E-06 1.118E+10 22 1.111E-01 3.061E+10 48 1.125E-06 8.577E+09 23 6.738E-02 2130E+10 49 6.826E-07 1.068h10 24 4.087 5 02 1.233E+10 50 4.140E-07 1.337h10 25 2.554E-02 1.700E+10 51 2.511E-07 1.327E+10 26 1.989E-02 8.614E+09 52 1.523E-07 1.253E+10 53 9.237E-08 3.779E+10 Note: Tabulated energy levels represent the upper energy in each group. - 6-24
TABLE 6-12 Comparison of Calculated and Measured Neutron Exposure Levels for - Vogtle Unit 1 Surveillance Capsules Y and U Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE Rate for Capsule Y Calculated Measured C/M M/C Ruence (E > 1.0 Mev) [n/cm2-sec) 1.122E+19 1.242E+19 0.903 1.107 Ruence (E > 0.1 Mev) [n/cm2-sec] 4.853E+19 5.000E+19 0.970 1.030 dpa 2.199E-02 2.258E-02 0.974 1.027 Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE Rate for Capsule U Calculated Measured C/M M/C Ruence (E > 1.0 Mev) [n/cm2-sec] 3.289E+18 3.437E+18 0.957 1.045 Ruence (E > 0.1 Mev) [n/cm2-sec] 1.410E+19 1.481E+19 0.952 1.051 dpa 6.414E-03 6.490E-03 0.988 1.012 e l l O 6-25
1 l l TABLE 6-13 I Neutron Exposure Projections at Key Locations on the Pressure Vessel Clad / Base Metal Interface 1 BEST ESTIMATE EXPOSURE (4.M EFPY) AT THE PRESSURE VESSEL INNER RADIUS l l O' 15 25' 30 35' 45* E > 1.0 1.802E+18 2.678E+18 3.155E+18 2.000E+18 2.622E+18 2.942E+18 E > 0.1 3.623E+18 5.453E+18 8.332E+18 7.2ME+18 7.127E+18 dpa 2.657E-03 3.930E-03 5.0ME-03 4.208E-03 4.443E-03 BEST ESTIMATE EXTRAPOLATION FLUX AT THE PRF3SURE VESSEL INhTR RADIUS 0_ 15 25 30 )_5
- 45 E > 1.0 1.231E+10 1.829E+10 2.155E+10 1.366E+10 1.791E+10 2.009E+10 E > 0.1 2.474E+10 3.724E+10 5.690E+10 4.920E+10 4.868E+10 dpa 1.815E-Il 2.684E-Il 3.418E-11 2.874E-Il 3.034E-Il BEST ESTIMATE EXPOSURE (16.0 EFPY) AT THE PRESSURE VESSEL INNER RADIUS
_0 15' 25' 30 35* 45 E > 1.0 6.216E+18 9.236E+18 1.088E+19 6.897E+18 9.043E+18 1.015E+19 E > 0.1 1.249E+19 1.880E+19 2.873E+19 2.484E+19 2.458E+19 dpa 9.162E-03 1.355E-02 1.726E-02 1.451E-02 1.532E-02 l BEST ESTIMATE EXPOSURE (32.0 EFPY) AT THE PRESSURE VESSEL INNTR RADIUS 0 15 25 30'. 35* 45 l ! E > 1.0 1.243E+19 1.847E+19 2.176E+19 1.379E+19 1.809E+19 2.029E+19 E > 0.1 2.499E+19 3.761E+19 5.746E+19 4.968E+19 4.916E+19 dpa 1.832E-02 2.710E-02 3.451E-02 2.902E-02 3.0ME 02 . l 6-26 i i
TABLE 6-14 Neutron Exposure Values at the 1/4T and 3/4T Imations of the Reactor Vessel Base Metal FLUENCE BASED ON E > 1.0 MeV SLOPE O 15' 25* 30 35* 45 16 EFPY FLUENCE SURFACE 6.216E+18 9.236E+18 1.088E+19 6.897E+18 9.043E+18 1.015E+19 1/4T 3.375E+18 4.997E+18 5.941E+18 4.938E+18 5.407E+18 3/4T 7.210E+17 1.053E+18 1.284Ei-18 1.058E+18 1.096E+18 32 EFPY FLUENCE SURFACE 1.243E+19 1.847E+19 2.176E+19 1.379E+19 1.809E+19 2.029E+19 1/4T 6.750E+18 9.994E+18 1.188E+19 9.875E+18 1.081E+19 3/4T 1.442E+18 2.106E+18 2.568E+18 2.116E+18 2.191E+18 FLUENCE BASED ON dpa SLOPE
~
0 15 25' 30 35 45 16 EFPY FLUENCE SURFACE 6.216E+18 9.236E+18 1.088E+19 6.897E+18 9.043E+18 1.015E+19 1/4T 3.922E+18 5.782E+18 7.072E+18 6.005E+18 6.473E+18 3/4T 1.361E+18 2.004E+18 2.600E+18 2.252E+18 2.212E+18 32 EFPY FLUENCE SURFACE 1.243E+19 1.847E+19 2.176E+19 1.379E+19 1.809E+19 2.029E+19 1/4T 7.844E+18 1.156E+19 1.414E+19 1.201E+19 1.295E+19 3/4T 2.722E+18 4.009E+18 5.201E+18 4.504E+18 4.423E+18 I ( 6-27
TABLE 6-15 Updated Lead Factors For Vogtle Unit 1 Surveillance Capsules CAPSULE LEAD FACTOR U WITHDRAWN EOC 1 i V 3.83
~
X 4.12 W 4.12 Y 3.83* Z 4.12
- WITHDRAWN EOC 4, BASIS FOR THIS ANALYSIS e
e 6-28
SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The surveillance capsule withdrawal schedule of Table 7-1 meets the requirements of ASTM E185-82 and is recommended for removal of future capsules from the Vogtle Unit I reactor vessel. TABLE 7-1 Vogtle Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluence Capsule Imation lead Factor (EFPY)") (n/cm2 , E>l.0 MeV) '
'U 58.5 4.01 1.14 3.44 x 10"*
Y 241.0* 3.83 4.64 1.24 x 10"N V 61.0 3.83 8.37 - 2.18 x 10"M l X 238.5* 4.12 11,64 3.26 x 10" i W 121.5* 4.12 Stand-by -- l_ Z 301.5* 4.12 Stand By -- l (a) Effective Full Power Years (EFPY) from plant startup. (b) Actual measured neutron fluence (c) Approximate EOL (32 EFPY) peak vessel inner surface fluence.
~
i l ( 7-1 l i
SECTION
8.0 REFERENCES
l' l
- 1. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.
Nuclear Regulatory Commission, May,1988. l
- 2. Singer, L.R., Georgia Power Company Alvin W. Vogtle Unit No.1 Reactor Vessel Radiation Surveillance Program, WCAP Il011. February 1986.
- 3. ASTM E185-82, Standard Practicefor Conducting Surveillance Testsfor Light-Water Cooled l Nuclear Power Reactor Vessels, E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
l
- 4. - Section III of the ASME Boiler and Pressure Vessel Code, Appendix G, Protection Against l Nonductile Failure.
- 5. ASTM E208, Standard Test Methodfor Conducting Drop-Weight Test to Determine Nil. Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3 American Society for Testing and Materials, Philadelphia, PA.
- 6. S.E. Yanicb.ko, S.L. Anderson, L. Albertin, N.K. Ray, Analysis of Capsule Ufrom the Georgia Power Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-12256, May 1989.
1
- 7. Code of Federal Regulations,10CFR50, Appendix G, Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 8. ASTM E23-92, Standard Test Methods for Notched Bar impact Testing of Metallic Materials, in l
ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1992. 8-1
i l l
- 9. ASTM A370-92, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993. -
- 10. ASTM E8 91, Standard Test Methods of Tension Testing of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1992.
- 11. ASTM E21-79(1988), Standard Practicefor Elevated Temperature Tension Tests of Metallic l
Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1992.
- 12. ASTM E83-92, Standard Practicefor Venfication and Classification of Extensometers, in AS'IM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.
- 13. R. G. Soltesz, R. K. Disney, J. Jedmch, and S. L. Ziegler, Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique, WANL-M(LL)-034, Vol. 5, August 1970.
- 14. ORNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Libraryfor Light Water Reactors.
- 15. R. E. Maerker, et al, Accounting for Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis, Nuclear Science and Engineering, Volume 94, Pages 291-308, 1986.
- 16. S. T. Lesho, W. R. Perry, et al., The Nuclear Design and Core Physics Characteristics of the Alvin W. Vogtle Unit i Nuclear Power Plant Cycle 1. Westinghouse WCAP-11338, November, 1986.
- 17. K. A. Potter, K. W. Bonadio, The Nuclear Design and Core Physics Characteristics of the Alvin -
W. Vogtle Unit 1 Nuclear Power Plant Cycle 2, Westinghouse WCAP-11980 October,1988.
~
82
- 18. K. A. Potter, K. W. Bonadio, et al., The Nuclear Design Reportfor the Vogtle Electric Generating Plant, Unit 1, Cycle 3. Westinghouse WCAP-12480, February,1990.
- 19. K. A. Potter, K. W. Bonadio, et al., The Nuclear Design Reportfor the Vogtle Electric Generating Plant, Unit 1, Cycle 4, Westinghouse WCAP-13023, September,1991.
- 20. K. A. Potter, K. W. Bonadio, et al., The Nuclear Design Reportfor the Vogtle Electric Generating Plant, Unit 1, Cycle 5, Westinghouse WCAP-13607, February,1993.
- 21. ASTM Designation E482-89, Standard Guidefor Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 22. ASTM Designation E560-84, Standard Recommended Practicefor Extrapolating Reactor Vessel Surveillance Dosimetry Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 23. ASTM Designation E693-79, Standard Practicefor Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 24. ASTM Designation E706-87, Standard Master Matrixfor Light-Water Reactor Pressure Vessel Surveillance Standard, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
l
- 25. ASTM Designation E853-87, Standard Practicefor Analysis and Interpretation of Light-Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991,
- 26. ASTM Designation E261-90, Standard Methodfor Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
8-3
- 27. ASTM Deslgnation E262-86. Standard Methodfor Measuring Thermal Neutron Flux by Radioactivation Techniques,in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991. .
- 28. ASTM Designation E263-88, Standard Methodfor Determining Fast-Neutron Flux Density by Radioactivation ofIron, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 29. ASTM Designation E264-87, Standard Methodfor Determining Fast-Neutron Flux Density by Radioactivation of Nickel, in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA,1991.
- 30. ASTM Designation E481-86, Standard Methodfor Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 31. ASTM Designation E523-87, Standard Methodfor Determining Fast-Neutron Flux Density by Radioactivation of Copper,in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 32. ASTM Designation E704-90, Standard Methodfor Measuring Reaction Rates by Radioactivation of Uranium-238, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 33. ASTM Designation E705-90, Standard Methodfor Measuring Fast-Neutron Flux Density by Radioactivation ofNeptunium-237, in ASTM Standards, Section 12, American Society for l Testing and Materials, Philadelphia, PA,1991.
I l
- 34. ASTM Designation E1005-84, Standard Methodfor Application and Analysis of Radiometric '
l Monitorsfor Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for > Testing and Materials, Philadelphia, PA,1991. -
\
l
- 35. F. A. Schmittroth FERRET Data Analysis Code, HEDL-TME 79-40, Hanford Engmeermg j Development Laboratory, Richland, WA, September 1979. l 8-4 ;
l
. . . =. - - . -_ .-. - . . .
- 36. W. N. McElroy, S. Berg and T. Crocket. A Computer-Automated Iterative Method ofNeutron Flux Spectra Determined by Foil Activation. AFWIeTR-7-41, Vol I-IV, Air Force Weapons ~ e Laboratory, Kirkland AFB, NM, July 1%7.
i
- 37. EPRI-NP-2188, Development and Demonstration of an Advanced Methodologyfor LWR Dosimetry Applications, R. E. Maerker, et al.,1981.
I \ l i i i l 1 l i l l l i
- l 8-5 i i
I l APPENDIX A Load-Time Records for Charpy Specimen Tests 9 1 1 I a e
5 Wp =
= Wg = =
Pg= M A XIMUM LOAD p y PF= FR ACTURE LOAO I PGY "Ob b b #I- , p I YlELO LOAD I I O I 4 O , I
> 1 I FAST' FRACTURE I
P4 = ARREST LOAD l l / I i . I l I I i <
$ l I i 1
l i i i l ! i I I I
->- tyg .
e tg r 4
-tp =
TI M E t gy = Time to general yielding wg = Fracture initiation region tg = Time to 2naximum load wp = Fracture propagation region ty - Time to fast (brittle) fracture start 4 Figure A-1. Idealized load-time record 4
-.-_4. _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ = c- ,
. . . . . . . ... . ~ . . ~ . = . . . . .. . - , . . . . . - . . , . ..
1 1 1 I
.I
, V0GTLE 91 "Y" AL67 g i s i l.
l
$4 e
5 a*- , w f. l
~_ _
4 l o ka ._
.D 1.2 2.4 3.6 4.8 6.0 TIME ( MSEC > ,
VOGTLE #1 "Y" AL67 : LONG : ! u L VDGTLE 01 "Y" AL69
- a a 8
. i "1 em_
a A 5 n*- v 4
=_
J o t L 1.2 2.4 e 3.6 i 4.8 i 6.0 l
.V TIME C PtSEC ) , - VOGTLE #1 "Y" AL69 : i LONG :
Figure A 2. Load-time records for Specimens AL67 and AL69 A-2 -
,, -y . , , , , . . , _ - . . . -,: , -~ . .
. - - . >- - - . . . . . _ _ . . . . . . _ _ . ..->m . . _ __. _ _ _ _ _ _ _ _ _ - _ _ . . _
?
, V0GTLE 01 "Y" AL74 i i a i 4
e- -
"1 4 .
r a 39- n v N a N-
.V 1.2 2.4 3.6 4.8 6.0 TIE C PtSEC >
VOGTLE #1 "Y" AL74 : LONG :
, VDGTLE 01 "Y" AL61 g a a
- e- -
1 4 a i 59- n v l e- _ a l 1
.V 1.2 2.4 3.6 4.8 6.0 TIE < PtSEC >
VOGTLE #1 "Y" - ' AL61 : LONG : . Figure A-3. Load-time records for Specimens AL74 and AL61 A-3 I
- - - . . . - - - ...- . - - - .. -- - - . .. ~- .. . - . . ~ . . - . . . . . .
b
, VDCTLE #1 "Y" AL64 a i .- g # e
.4 1 m_ 4 A 39- a
. v w
a j
*f. -
o s i 6.0 3.6 i 2.4 4.8
.0 .1. 2 TIE < MSEC )
VOGTLE #1 "Y" AL64 : LONG :
- vocTLc s1*Y" AL65
, i i
g i 9_ a4 a S n9-l v, _ N i 3 i l l \_ ' ) s n o s 2.4 s 3.6 4.8 6.0
.0 1.2 TIME C MSEC )
VOGTLE #1"Y" AL65 : !
- LONG : 1 Figure A-4. Load-time records for Specimens AL65 and AL64 I
9 A4 e I 1.. . l
! 4
~ . _ , _ . . _. _ . . _ . .
vocTu: st*Y" AL62 i i e 9 . . a a 1 =;_ . 7 n
~
S *M-- , w N i 9-l --_-.... - n 1.a' .e.4' 3.6' 4.s' 6.o ! TIrc ( nste ) VOGTLE #1"Y" AL62 : LONG : YDGTLE s1 "Y" AL71
- i a 8 .
e a - 1_ I a 3 M*- v a , q_ ' l ~- a i
.. . i 6.0
.D 1.2 2.4 3.6 4.8 TITE < ttSEC > !
VOGTLE #1 "Y" - AL71 : LONG : 4 Figure A-5. Load-time records for Specimens AL62 and AL71 A-5 F"*' -v- - --,v,-,.mmm -s ow - ee- w- y.y-yr- w
, V?GTLE #1 "Y" AL73 i s s e
4 O m_ 54
~
m A e g_ a v q_ s o a 1 a 6.0
.c 1.2 2.4 3.6 4.8 TIMC ( MSEC )
VOGTLE #1 "Y" AL73 : LONG : VDGTLE 01"Y" AL72
. , I 4 i i
; m_ -
S n9-v n- - o . . i .
.D 1.2 2.4 3.6 4.8 6.0 Tite < MSEC )
VOGTLE #1"Y"
~
AL72 : LONG : Figure A-6. Load 4ime records for Specimens AL73 and AL72 A4
, VOCTLE #1 "Y" AL75 i i i g .
9_ _
&4 - 4 I !
n S n*- v N l N_ I o e i i i '
.D 1.2 2.4 3. 6 4.8 6.0 TIME ( MSEC )
VOGTLE #1 "Y" AL75 : LONG : i
, vocTLc e1 "Y" AL66 ,
i i i j i 9_ .. a4 7
^ i 1
3._ e v w_ N N_ o , a s
.V 1.2 2.4 3.6 4.8 6.0 TIME ( MSEC )
~
VOGTLE #1 "Y" AL66 : LONG : , Figure A-7. Load-time records for Specimens AL75 and AL66 A-7 l
l t l l l i l l i l 1
, VDCTLE 81 "Y" AL68 l i i i 5 I*
i* 54 7
^
l - m d v N
~_ -
1 a , a i i
.D 1.2 2.4 3.6 4.8 6. 0 TIME < MSEC )
VOGTLE #1 "Y" AL68 : LONG :
. VOCTLE 81 "Y" AL63 i i J 3 ,
7 m_ - a4 7 a
~
$ *- n v
m 1
- i i , ,
.D 1.2 2.4 3.6 4.0 6. 0 TIME ( M3EC ) ,
VOGTLE #1 "Y" l AL63 : l LONG : ) Figure A-8. Load-time records for Specimens AL68 and AL63 j A-B l l 1
V "Y" AL70
*- 0GTLE 81 . . i i O m-a4 .
7
^
l 9 *"- . I v I, .J q-o i i i
.0 1.2 i
2.4 3.6 4.8 6.0 TIME C PtSEC > VOGTLE #1 "Y" AL70 : LONG :
, VDCTLE 01"Y" AT65 i i g i a ? e- -
a4 E a 9 *-M v q- - I A ^_a_ m o a i i
.% 1.2 2.4 3.6 4.8 6. 0 i TINC ( PCEC >
VOGTLE #1"Y" -!) AT65 : TRANS : . 1 I Figure A-9. Load-time records for Specimens AL70 and AT65 A-9
I I I l
\
l V0GTLE #1"Ya AT64 i i l
". i j
. e
?_
S4 7 e
- 1 g ._ l m
v
.2
'f. -
j l l I l _ _ _ _ ^ ^
- 6.0 i
o i 3.6 i 2.4 4.8
.D 1.2 TIME < MSEC >
VOGTLE #1"Y" AT64 : TRANS :
. , VDCTLE 91 "Y" AT66
. i e i i 7 e_ _
h e e _ ci v
.J
.'" . - 1 k___ .
i
, i
)
og 1.2 2.4 3.6 4.8 6. O ; TIMC ( MSEC ) I VOGTLE #1 "Y" AT66 :
. TRANS :
Figure A-10. Load-time records for Specimens AT64 and AT66 . A.10 l 1 i
I i i l I 1 . i 1 i l I AT72 record not available ; due to computer malfunction l l vocnr et y- ar7s
; i i i i ! 5- -
E S n*- - v al N- - f n I s s , 1
.0 1.2 2.4 3.6 4.8 6. 0 i I
TIME ( PISEC > VOGTLE #1 "Y" AT75 : TRANS : 1 Figure A-11. Load time records for Specimens AT72 and AT75 A-11
. _ . . - _ .. . .. -. . - - - ~ ~ _ _ __ _ . - . .
t l l
-VOCTLE #1"Y" AT70 '1
- i a i e i
n
- 1e, i a
5 *- " i v e um l '.~ N i
*f l
l.- -:~__ - - __ _ _ - i a o i 4.8 6. 0
,D 1.2 i
2.4 3.6 TIME < MSEC > VOGTLE #1"Y" t AT70 : TRANS : l i V0GTLE #1 'Y' AT68 !', , e a s e 4 9._
&4
~
m A !
= -
l
$ *- " i l
l ga
~'
i I i
*f
~
i l i i i o i e 4.8 6. 0
.D 1.2 2.4 3.6 l
TIME ( MSEC ) VOGTLE #1 "Y" i, AT68 .: TRANS : 1 Figure A-12. Load-time records for Specimens AT70 and AT68 A-12 1 i ,
YOCTLI #1'Ya AT63 l i i i j )
-l l
0_ h# l ^ { $ $_ n i v y_ e . ~~ g - _ og 1.2 2.4 3.6 4.8 6.0 TIPE < MSEC ) VOGTLE #1"Y" AT63 : TRANS :
, YOCTLE N1 "Y" AT73 s e I s 9_ _
&4 7 a 9 g_ _ n v q_ _ o e a s s
.D 1.2 2.4 3.6 4.8 6. 0 TIMC ( MSEC )
VOGTLE #1 "Y" AT73 : TRANS : Figure A-13. Load: time records for Specimens AT63 and AT73 A 13
_ . . _ . _..m. . . . . . _ . _ _ . . . _ _ . . , _ _. . . - . . . . . . , . . _ _ . .._..m. . _. ._ __
..~_..m..,
l
, VOCTLE 01 "Y" AT69 i i i g i "1 4. _ ~ .
m i a
- 4 53- a !
v
- i 1
0 q_ I t
= , i.
.V 1.2 2.4 3.6 4.8 6.0 TIE < MSEC )
~VOGTLE #1 "Y" AT69 :
' TRANS :
. . YOGTLE 01 'Y" AT71 g i i s :
7._ - 84 m A
$ m 9-w N
9- _ f a i . i i
.0 1.2 2.4 3.6 4.8 6.0 TIE < MSEC )
VOGTLE #1 "Y" AT71 :
. TRANS :
Figure A-14. Load-time records for Specimens AT69 and AT71 1 i i l i 1 A-14
t i i i I
, VOCTLE S1 "Y" AT74 4 5 i I e- - ,
."1 ; .
z 59- n v ai y- - l
.9 1.2 2. 4 3. 6 4.8 6.0 TIPE < MSEC ) ,
VOGTLE #1 "Y" AT74 : TRANS :
, VDGTLE #1 'Y' AT67 j i i n .s.
j $- - E a S n9- - v af - _
- i e i .
.9 1.2 2.4 3.6 4.8 6. 0 TIME < MSEC >
VOGTLE #1 "Y" AT67 : TRANS : Figure A-15. Load. time records for Specimens AT74 and AT67 A 15
l VDCTLE G1 'Y' AT61 l , i i l g i a 1 l ?e _
- i. 54 1 E l a
3 *- n V l . Qd S + y_ -
- .9 1.2 2.4 3.6 4.8 6. 0
' TIME < PCEC > vogtle #1 "y" AT61 l l TRANS :
, VDCTLE at "Y" AT62
. i a i i e
&4 7
a 3*~ e v T em 1 I
.3 )
1 i N-
~ l fr N e a , s s
.D 1.2 2.4 3.6 4.8 6.0 TIME ( MSEC )
VOGTLE #1 "Y" AT62 : TRANS : Figure A-16. Load-time records for Specimens AT61 and AT62 l i l ' A-16 l
1 l l
, V!r.TLE e1 *Y* Au66 g 3 i i i O e_ -
a4 ' 1?
~
59- Y. _ I e v
*- _ l
*f. - _
- i i a 1.2 i
*V 2.4 3.6 4.8 6.O TIPE ( MSEC >
VOGTLE #1 "Y" AW66 : WELD :
, VDGTLE 01'Y" AW68 e i a i .
h
~
39- , v A
=_ -
e i i a i
.g 1.2 2.4 3.6 4.8 6.0 TIME < MSEC )
VOGTLE #1"Y" AW68 : WELD : Figure A-17. Load-time records for Specimens AW66 and AW68 A-17
i I l
)
, VOCTLE #1 "Y" aW65
. a a i s ,
4 a _ ~ l ._ s l E n j S o*- 4 i I w l
~
s-l . P u_
.9 1.2 2.4 3.6 4.8 6.0 TIE < MSEC )
VOGTLE #1 "Y" AW65 : WELD : e vocTLE et y- aus4
. , s s e
. a e
i
~
a I m_ E n S*- a v ' v, _ N
=_
a 2.4- 3.6 4.8 6.0
.5 12 TIE ( MSEC >
VOGTLE #1"Y" AW74 : WELD : ; 4 Figure A.18. Load-time records for Specimens AW65 and AW74 A.18 A y ..- , i..,-.,.,,_,,. c._ ,_ , . . _ , , ._,y.. .. . . , . . - -
. . . ~ ~ . - = . ~ . ... . . . . . . - - - - - - . .~.
I i
, VOCTLE 01 'Y' AW62 s n a s 4
e- -
"1 4 .
^
l s 9- n i w
\
l - . i 1 <- - l k'3 i t u- - f-o . i a
.9 1.2 2.4 3.6 4.8 6. 0 l TIME' ( MSEC )-
l VOGTLE #1 "Y" AW62 : ~ WELD :
. l
- vocTLE e!"Y" AW67
, o e a .
4 , i r e-
"1 4 E
a l - - 3 g_ n l w { w- l 3
- l N-l l
f -
, = , - - -
.t 1.2 2. 4 3.6 4.8 6.0 '
TDC ( MSEC ) VOGTLE #1"Y" . , AW67 : ) WELD 1
. Figure A-19. Load-time records for Specimens AW62 and AW67 e
A-19
. . . .._ . . _ . . _ . . . - . . _ . . .m..,_ . . _ - .. _.. _
i I i I: ! , YDGTLE si "Y' Au61 g i i i i
, 1" ;e Y
1 1 S "*- 1 v 4 s-QA
. 3 i
q-o e a i i
.0 1.2 2.4 3.6 4.8 6.0 TIE < MSEC )
VOGTLE #1 "Y" i
-AW61 :
WELD :
, V0GTLE 01 "Y" AW63 j i i a a t
"I we I
a 3 g- - n v v A n- - N e i i i
.9 1.2 2.4 3.6 4.8 6. 0 TIE ( MSEC )
VOGTLE #1 "Y" AW63 : WELD :
. Figure A-20. Loat. Lme record.s for Specimens AW61 and AW63 A.20
, YDCTLE #1 "Y" Au70 g i i e i ? e- -
g4 - E a 39-m v N N- -
.0 1.2 2.4 3.6 4.8 6.0 TIME ( MSEC )
VOGTLE #1 "Y" AW70 : WELD :
, V0GTLE #1 "Y" AW72 i i i j i ? e- -
L4 7 a 59- m v N N- - l o i i i i
.t 1.2 2.4 3.6 4.8 6.0 TIME ( MSEC )
VOGTLE #1 "Y" ' i AW72 : WELD : Figure A-21. Load-time records for Specimens AW70 and AW72 ! A-21
l l l l
, VDCTLE #1*Y" AW71
. s i s e
O e_ -
. &4 1 1 I
s l 1
$ *M-v
- a l
=_*
o a i a i
.D 1.2 2.4 3.6 4.8 6. O TIRE ( MSEC >
VOGTLE #1"Y" AW71 : WELD : V0GTLE #1*Y" AW69 l- .- g i i e a 7 e_ _ a4 m a
$*- M v
N u_ - 1 o i i e i
.D 1.2 2.4 3.6 4.8 6. 0 i TIME < MSEC ) l l
VOGTLE #1"Y" l AW69 : WELD : I
- l l
t j Figure A.22. Load time records for Specimens AW71 and AW69 - l l l . i l [ A 22 ' I i
..., , u ... -- .. . . - .-.
l 1-
, VDGTLE 01 "Y" AW75
; e "e-14 1
^
i 39- n -
- l i
3- - i QN 5 . u- - i i 1. l .c .1.2 2.4 3.6 4.s 6.o l titc - < nsEc > l VOGTLE #1 "Y"' l AW75 : WELD : 1 i
, V0GTLE 01 "Y" AW73 g i i a a "I ee-a a g- -
n v
-l
~
2~
]
l n- - o . . . .
.D 1.2 2.4 3.6 4.8 6.0 ;
Tite ( MSEC )- ' VOGTLE #1 "Y" - l Batch:AW73 ! 4 Figure A-23. Load-time records for Specimens AW75 and AW73 i l l A-23
, y --- er ,,mm-- - s - - , . . . , . - , . - - . . . , , . - ,
l t I l t VOCTLE el "Y' Au64 ;
, l l
7 m_ l 1- 54 7 n t
$ n9-I
- v. -
N m_ o i i 6.0 0 1.2 2.4 3.6 4.8 TIPE < PtSEC ) VOGTLE #1 "Y" AW64 : WELD : YDCTLE #1 'Y" AH75 4 7 m_ _ E4 7 a S M9-v N u_ I e a >
.0 1.2 2.4 3.6 4.8 6.0 TIPE ( PtSEC )
VOGTLE #1 "Y" Batch:AH75 Figure A-24. Ioad time records for Specimens AW64 and AH75 A 24
_ _ . _ _ _ _ .~ ._m . . . . . _ - . ~- _ .. .. .-, .
, VDCTLE 01 "Y" M468 4 4 4 3
m-
"k 4 -
E n 51- n v l , u u- - o , , s ,
.D 1.2 R.4 3.6 4.8 6.0 TIE C MSEC )
VOGTLE #1 "Y" Batch:AH68 l l l
, VOCTLE 01*Y" AH63 i
g i s , m- -
"1 4 i A
59- n v I u- - i O cm , i i . f .D 1.2 2.4 3.6 4.8 6.0 TIE ( MSEC >
.VOGTLE #1"Y" AH63 :
HAZ : . Figure A-P.5. Load-time records for Specimens AH68 and AH63 A-25
l VDGTLE 01 "Y" AH67
, i l- i i i g
~
? e_
&4 7
n S .n 9-t v N l ! N_ a.; s.6' <.s' 6.o TIME ( MSEC ) VOGTLE #1 "Y" AH67 HAZ
, VDGTLE #1 "Y" AH72 g i i i i 7_ _
h a 3 n$_ _ v v_ J N_ o i . .
.D 1.2 2.- 3.6 4.8 6.0 TIME ( MSEC )
VOGTLE #1 "Y" AH72 : HAZ : Figure A-26. Load-time records for Specimens AH67 and AH72 A 26
vocTt.c es v' aH56 i i 4 . i W . a 1_ 4 7 n
$ *e-v Ed a
N-l l 3.6' 4.s'- 6.0 l , i.e ' a.4'- l
. TIE < MSEC >
VOGTLE #1"Y" AH66 HAZ : l ! . voctut et av- act- , , '
; i . . .
i i j 5- - 7
^ l 5 j- .
w i ! N-i
.D 1.2 2. 4 3.6 4.8 6. 0 TIE ( MSEC )
VOGTLE #1 "Y" AH71 : HAZ : , j
. Figure A 27. Load-time records for Specimens AH66 and AH71 ll A 27 i l l I l , - _ . _ -- .- . . ._..._._.a
. _ . . ._ ._ _.. . _ . _ _ m . . _ _ _ . _ . _ _ . . , _ . . . . . . . __. ..--_.m_ m . . . .-,. .... _ _ .. . .
l l l l
- V0GTLE #1 "Y" AH74 l i i i i
" e- -
- f. I J V
n 5 *n-v i ,_ s4 ; 9 1 u_ l I l 1. , ' . 4' 3. 4. 6.o TIE ( ttSEC ) l VOGTLE #1 "Y" ! AH74 : l FAZ :
, vocTtt et v- aH6s g i i i i i e- _
& v 7
a 3 ._ n w e_ t
=_ _
}
o . i 7- ,
.9 1.2 2.4 3.6 4.8 6.0 TIE ( ftSEC )
VOGTLE #1 "Y" , AH65 : HAZ : 1 Figure A-28.- Load-time records for Specimens AH74 and AH65 A-28
,.. r m e -.- --3 y ..y---. w , 4 ---+- r- me. --
W w + +*r =e+ ',e*-
VOCTLE 01"Y" AH61
, i a e
g i
~
a - I; ._s - . a
$ *a-w el u-f _
. , c ,
1.2 2.4 3.6 4.8 6.0
.D TIPE ( MSEC )
VOGTLE #1"Y" AH61 : HAZ : VOCTLE el*Y' AH73 i s a g i , a _ 1._ 4 . I e m v s-u u- - o i i i .
.D 1.2 2.4 3.6 4.9 6.0 T!!C ( MSCC >
VOGTLE #1"Y" . AH73 : HAZ : Figure A-29. Imad-time records for Specimens AH61 and AH73 A 29
, V0GTLE #1 "Y" AH69 .
i i i j l 7 co _ _ 54 7
^
3 N._ l i w ' 1 N m_ _ 1 o i i i i
.D 1.2 2.4 3.6 4.8 6. 0 TIME < MSEC )
VOGTLE #1 "Y" AH69 : HAZ :
, VDCTLE 41 "Y' AH62 g i a i i 9 e_ _
a4 7 a 3 $_ n v a l m_ _ o e i a
- i
.D 1.2 2.4 3.6 4.8 6.0 TIME < MSEC >
VOGTLE #1 "Y" t AH62 : ! HAZ : Figure A-30. Load-time records for Specimens AH69 and AH62 i l A-30 l l
, YDGTLE 01 "Y" AH70 i i g i i ? e- -
54 ' 7 s S 9-n v I 1 as - o i , a i
.D 1.2 2.4 3.6 4.8 6. 0 TIPE < ttSEC )
VOgtle #1 "y" AH70 : ' HAZ :
, VOCTLE 01 "Y* AH64 i a i i j ,
9 e- - E4 n S n9- - v w- - u- - o , i i i I 2.4 3.6
.D 1.2 4.0 6. 0 TIME ( PtSEC )
VOGTLE #1 "Y" . ; AH64 : l HAZ : 1 i Figure A-31. Load time records for Specimens AH70 and AH64 l A-31 l l
l ! l l
. APPENDIX B l HEAYUP AND COOLDOWN LIMIT CURVES 1
FOR NORMAL OPERATION OF THE VOGTLE UNIT 1 REACTOR PRESSURE VESSEL i l
~
i 4 4 9 e
- m
l i l TABLE OF CONTENTS 1 Section Title Page . i LIST OF ILLUSTRATIONS B-2 i ! LIST OF TABLES B-2 ! I - l l B-1. INTRODUCTION B-4 i B-2. FRACTURE TOUGHNESS PROPERTIES B-5 B-3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS B-5 l B-4. CALCULATION OF ADJUSTED REFERENCE TEMPERATURE B-8 B-5. HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES B-9 , 1 B-6. REFERENCES B-22 l l i l l t B-1
LIST OF ILLUSTRATIONS Figure, Title Pace B-1 Vogtle Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates up to B 20 100*F/hr) Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors, and Margin of 74 psig for Pressure Difference
~
Between Pressure Instrumentation and Reactor Vessel Beltline Region) Including 10% Increase in Pressure for Temperatures Less than 200 F per AShG Code Case N-514 B-2 Vogtle Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates B-21 up to 100*F/hr) Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Enors, and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor Vessel Beltline Region) Including 10% Increase in Pressure for Temperatures less than 200 F per ASME Code Case N-514 LIST OF TABLES Table Title Page I B-1 Vogtle Unit 1 Reactor Vessel Toughness Table (Unirradiated) B-12 l l B-2 Calculation of Average Copper and Nickle Weight Percent For Intermediate B-13 Shell Plates Using All Vogtle Unit 1 Chemistry Test Results B-3 Calculation of Average Copper and Nickle Weight Percent For Lower B-14 Shell Plates Using All Vogtle Unit 1 Chemistry Test Results B-4 Calculation of Average Copper and Nickle Weight Percent For Weld B 15 Metal Using All Vogtle Unit 1 Chemistry Test Results I B-2 i I
LIST OF TABLES (cont.) l Table Title Page , , B-5 Calculation of Chemistry Factors Using Surveillance Capsule Data B-16 1 l B-6 Summary of Adjusted Reference Temperatures (ART's) at 1/4-t and 3/4-t B-17 f - Imations for 16 EFPY B-7 Calculation of Adjusted Reference Temperatures at 16 EFPY for the Limiting B-18 l l Vogtle Unit 1 Reactor Vessel Material - Intermediate Shell Plate B8805-2 B-8 Vogtle Unit 1 Heatup and Cooldown Data at 16 EFPY with Margins of 10'F B-19 and 60 psig for Instmmentation Enors i I 9 l i t 1 ( l i B-3
l l B-1. INTRODUCTION l Heatup and cooldown limit curves are calculated using the most limiting value of RT 3or (reference nil-ductility transition temperature) corresponding to the limiting beltline region material for the reactor vessel. The most limiting RTyor of the material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties and estimating the radiation-induced ARTxo7 The unirradiated RTxor is designated as the higher of either the drop weight nil-ductility transition temperature (NDTI') or the temperature at which the material exhibits at I least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F. I RTyor increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTxor at any time period in the reactor's life, ARTsar due to the radiation exposure associated l with that time period must be added to the original unirradiated RTyor. The extent of the shift in RTsor is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 (Radiation Embrittlement of Reactor Vessel Materials)t"R Regulatory Guide 1.99, Revision 2 is used for the calculation of adjusted reference temperature (ART) values (irradiated RTso7 with margins for uncertamties) at 1/4-t and 3/4-t locations.
"t" is the thickness of the vessel at the beltline region measured from the clad / base metal interface.
The pressure-temperature limit curves in Figures B 1 and B-2 of this report include margins of 10 F and 60 psig for instrumentation errors and include a pressure adjustment of 74 psig for the pressure t l difference between the wide-range pressure transmitter and the limiting reactor vessel pressure in the beltline region. In addition, the pressure-temperature limit curves contain the 10% relaxation in l pressure for temperatures less than 200 F, per ASME Code Case N-514". The application of ASME I I Code Case N-514 increams 'be operating margin in the region of the pressure-temperature limit curves where the low temperature overpressure protection (LTOP) system is active. Although expected soon, use of the Code Case N-514 has not yet been formally approved by the NRC and therefore it is recommended that Georgia Power Company interface with the NRC to obtain their formal approval on
, this application of ASME Code Case N-514.
~
l B-4 l l
B-2. FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materialin the reactor coolant pressure boundary are - determined in accordance with the NRC Regulatory Standard Review Plan ts2) The pre-irradiation fracture-toughness propenies of the Vogtle Unit I reactor vessel are presented in Table B 1. The irradiated fracture toughness properties of the reactor vessel beltline material were obtaired directly j from the Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program. Credible surveillance data, as l defined in Regulatory Guide 1.99 Rev. 2, is available from capsules U and Y. I B-3. CRt i . AA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ! l l The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K ,i for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Km, for the metal temperature at that time. Ka is obtained from the reference fracture toughness curve, defined in 1 Appendix G of the ASME Code'881 The Km curve is given by the following equation: l l I l Km = 26.78 + 1.223
- e " *"" + "*1 (1) where, t
I l Km= reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility transition temperature RTNDT ; 1 I The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME l t Code *'l as follows: I i C*Km+Krr <_ Km (2)
\
l where, l Km = stress intensity factor caused by membrane (pressure) stress i i Krr = stress intensity factor caused by the thermal gradients 1 B-5 1
)
l Km = function of temperature relative to the RTm of the material
. C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions when the reactor core is not critical At any time during the heatup or cooldown transient, Km is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTm, and the reference fracture toughness curve.
The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Krr, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall.
- - During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside that increase with increasing cooldown rates.
Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4-t vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor l coolant temperature, the temperature difference across the wall developed during cooldown results in a l higher value of Km at the 1/4-t location for finite cooldown rates than for steady-state operation.
, Furthermore, if conditions exist so that the increase in Km exceeds Krr, the calculated allowable l l
pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4-t location I i B-6 l l l l l l
l l and, therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire moldown period. . Three separate calculations are required to determine the limit cmves far finite heatup rates. As is I done in the cooldown analysis, allowable pressure-temperature relationahips are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4-t defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature: therefore, the Km for the 1/4-t crack during heatup is lower than the Km for the 1/4-t crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K m 's do not offset each other, and the pressure-temperature curve based on steady state conditions no longer represents a lower bound of all similar cmves for finite heatup rates when the 1/4-t flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained, l The third portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4-t deep outside su. face flaw is assumed. Unlike the situation at the vessel l inside surface, the thermal gradients established at the outside surface during heatup produce stresses l which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by constmeting a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations . because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times - be based on analysis of the most critical situation. B-7
Finally, the 1983 Amendment to 10CFR50" has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the
. closure flange regions must exceed the material unitradiated RTxm by at least 120*F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Vogtle Unit 1).
i Table B-1 Indicates that the limiting unirradiated RTym of 20'F occurs in the closure head flange of the Vogtle Unit I reactor vessel, so the minimum allowable temperature of this region is 150*F at press. es greater than 561 psig. These values include margins of 10*F and 60 psig for instmmentation errors. This limit is shown in Figures B 1 and B-2 wherever applicable. i B-4. CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2", the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression: ! ART = RTNDT = Initial RTsm + ARTym + Margin (3)
- Initial RTym is the reference temperature for the unirradiated material as dermed in paragraph NB-2331 of Section III of the AShE Boiler and Pressure Vessel Code. If measured values of initial RTym for the material in question are not available, generic mean values for that class of material are l used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTxm is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows 1 l l ART 3m = CF
- f$2s . nio ios o (4)
CF (*F) is the chemistry factor, obtained from the tables in Reference B1, using the average values of the copper and nickel content as reported in Tables B-1 through B-4. Some of the chemistry factors were also calculated using the surveillance capsule data in Table B-5. l To calculate ARTym at any depth (e.g., at 1/4-t or 3/4-t), the following formula is used to attenuate the i fluence at the specific depth. B-8 l
f, = f,s ,* e *'* (5) where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal . Interface. The resultant fluence is then put into equation (4) to calculate ARTxo1 at the specific depth. The calculated peak surface fluence for Vogtle Unit 1 intermediate and lower shell plates and 2 circumferential weld at 16 EFPY is 1.088 x 10" n/cm (E>1.0 MeV) as indicated in Section 6.0 of this 2 report. The surface fluence for the longitudinal welds at 16 EFPY is 6.216 x 10" n/cm (E>l.0 MeV) for weld seams located at 0* and 180 , and 6.897 x 10" n/cm2 (E>1.0 MeV) for weld seams located at 30'. All materials in the beltline region of Vogtle Unit I reactor vessel were considered in determining the limiting material. The adjusted reference temperatures at 1/4-t and 3/4-t are summarized in Table B-6. From Table B-6, it can be seen that the limiting material is the Intermediate Shell Plate B8805-2 for heatup and cooldown curves applicable up to 16 EFPY. Sample calculations to determine the ART values for intermediate shell plate B8805-2 at 16 EFPY are shown in Table B-7. B 5. HEATUP AND COOLDOWN PRESSURE TEMPERATURE LIMIT CURVES i Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections B-3 and B-4 of this reportI80 Since indication of reactor vessel beltline pressure is not available on the plant, the pressure difference between the wide-range pressure transmitter and the limiting beltline region must be considered. Generic calculations have determined that the pressure indicated by the reactor coolant system wide-range instrumentation should be assumed to be 74 psig less than that at the reactor vessel beltline for Vogtle Unit lis24 The limit
)
curves presented in Figures B-1 and B 2 include this pressure difference. In addition, at the request of Georgia Power Company, the heatup and cooldown curves herein were adjusted to incorporate a 10% increase in allowable pressure for temperatures less than 200 F, per ASME Code Case N-514. Code Case N-514 also requires LTOP systems to be effective at coolant temperatures less than 200*F or at coolant temperatures corresponding to a reactor vessel metal temperature less than 1/4-t RTsor + 50'F, whichever is greater. This code case has been approved by ASME but has not yet been formally , accepted by the NRC. Figure B-1 presents the heatup curves with margins of 10T and 60 psig for possible instrumentation B-9
errors using a heatup rates up to 100*F/hr applicable for the first 16 EFPY. Figure B-2 presents the cooldown curves with margins of 10*F and 60 psig for possible instrumentation errors using cooldown
. rates up to 100*F/hr applicable for the first 16 EFPY, Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures B 1 and B 2. This is in addition to other criteria that must be met before the reactor is made critical.
I The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure B-1. The straight-line ponion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10CFR Part 50". The governing equation for the hydrostatic test is defined in Appendix G to Section III of the ASME Code as follows: 1.5 Kai :s Km (6) where, Km is the stress intensity factor covered by membrane (pressure) stress, Km = 26.78 + 1.233 e**8 " * "" " ",
~
T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility transition temperature. The criticality limit curve shown in Figure B-1 specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference B4. The pressure-temperature limits for core operation (except for low power physics tests) require that the reactor vessel be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section B-3. The minimum temperature for the inservice hydrostatic leak test for the Vogtle Unit I reactor vessel at 16 EFPY is 246*F. A vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. This curve includes margins of 10*F and 60 psig for
, instrumentation errors.
Figures B-1 and B-2 define all of the above limits for ensuring prevention of nonductile failure for the i Vogtle Unit I reactor vessel. These figures include an instantaneous pressure change at 200 F, as a B 10
result of applying ASME Code Case N-514, which could cause an additional pressure transient if the pressure-temperature limit curves were exactly followed. This transient can be avoided during normal heatup conditions by not exceeding the lower bound pressure value at 200 F until the reactor coolant , system water temperature is greater than 200*F (i.e., using the 60 F/Hr heatup curve in Figure B-1, 4 hold the pressure value of 837.97 psig constant for the temperature range between 189 F and 200 F). 1 For normal cooldown conditions, the curves allow an increase in allowable pressure at temperatures l below 200 F, therefore the pressure transient increases operating window margins and causes no adverse effect on cooldown operations. The data points used to develop the heatup and cooldown , pressure-temperature limit curves shown in Figures B-1 and B 2 are presented in Table B-8. i 1 i I i. l
\ ! l l
4 4 v l B-11
Table B-1 Vogtle Unit 1 Reactor Vessel Toughness Table (Unitradiated)I871 Material Description Cu (%)
- Ni (%)
- Initial RTxor (*F)
- Closure Head Flange --
0.70 20 *' Vessel Flange -- 0.71 0 *) Intermediate Shell Plate B8805-1 0.083 0.597 0 Intermediate Shell Plate B8805-2 0.083 0.610 20 Intermediate Shell Plate B8805-3 0.062 0.598 30 Lower Shell Plate B8606-1 0.053 0.593 20 Lower Shell Plate B8606-2 0.057 0.600 20 i Lower Shell Plate B8606-3 0.067 0.623 10 Intermediate & Lower Shell Vertical 0.039 0.102 -80
- Weld Seams and Girth Seam Weld i
(a) Initial RTsm values are measured values. t (b) These values are used for considering flange requirements for the beatup/cooldown curves". 4
- The average values of copper and nickel content dete mmed as indicated in Tables B-2 through B-4 on the following pages.
i i i . l 9 B-12
I 1 . Table B-2 I l Calculation of Average Copper and Nickle Weight Percent For Intermediate Shell Plates l Using All Vogtle Unit 1 Chemistry Test Results l l Inter. Shell Inter. Shell Inter. Shell 4 l Plate B8805-1 Plate B8805-2 Plate B8805-3 l Cu Ni Cu Ni Cu Ni l Reference (wt%) (wt%) (wt%) (wt%) (wt%) (wi%) . \ l l Matl. Cert. Repod883 0.09 0.60 l - l Matl. Cert. Repod88) 0.08 0.60
\
l l Chemical Analysis5 'l 0.08 0.59 Surveillance Program!"'1 0.08' O.59' Matl. Cert. Report
- 1 0.09 0.64
! Matl. Cert. Report"' 1 0.08 0.60 Chemical Analysistani 0.08 0.59 Surveillance Program5 'l 0.08' O.59' ! Matl. Cert. Repods 2i 0.07 0.60 Matl. Cert. Report ta 21 0.07 0.61 Chemical Analysis 5 ") 0.06 0.60 Surveillance Program5 'l 0.058 0.61 Surveillance Program5 'l 0.06' O.60* Capsule U Reportf871 0.053 0.586 i Capsule U ReportS 73 0.06* 0.60* l l Capsule U Reportf87 0.058' O.61* I l Chemical Analysis5 "3 0.061 0.584 l [ Average 0.083 0.597 0.083 0.610 0.062 0.598 l Not used in average calculation since same as in other references: reported only for canpleteness. l i 1 i
. j B-13 l
Table B-3 Calculation of Average Copper and Nickle Weight Percent For Lower Shell Plates Using All Vogtle Unit 1 Chemistry Test Results Lower Shell Lower Shell Lower Shell Plate B8606-1 Plate B8606-2 Plate B8606-3 Cu Ni Cu Ni Cu Ni Reference (wt%) (wt%) (wt%) (wt%) (wt%) (wt%) Matl. Cert. Repod8"1 0.06 0.61 Mad. Cert. Repodsui 0.05 0.58 Nr
- Chemical Analysista 'l 0.05 0.59 Surveillance Program f861 0.05* 0.59' Matl. Cert. Reportl82'l 0.07 0.64 Matl. Cert. Repod8"1 0.05 0.58 Chemical AnalysisI8"1 0.05 0.58 Surveillance ProgramI8 'l 0.05* 0.58' Mail. Cert. Repod" 'l 0.07 0.60 Matl. Cert. Report **'l 0.07 0.63 t
Chemical Analysis""1 0.06 0.64 Surveillance ProgramI8 'l 0.06* 0.64* Average 0.053 0.593 0.057 0.600 0.067 0.623 Not used in average calculation since same as in other references; reported only for completeness. l 1 I l I 1 B-14 i
Table B-4 l l Calculation of Average Coper and Nickle Weight Percent For Weld Metal I Using All Vogtle Unit 1 Chemistry Test Results . Weld Metal" Cu Ni Reference (wt%) (wt%) Surveillance Program I8 'l 0.030 -- - t8 Surveillance Program 'l 0.040 0.100 Surveillance Program!8'I 0.037 0.100 Capsule U Report *3 0.035 0.091 l Chemical Analysisf 8 2') 0.048 0.101 Chemical Analysist8 l 0.040 0.117 Chemical Analysistei4) 0.041 0.105 Chemical Analysis ts2n 0.040' O.100* Chemical Analysists22) 0.030* -- Average 0.039 0.102 . Not used in average calculation since same as in other references: reported only for completeness.
" The smvei!!ance weld is identical to that used in the core region for the longitudinal seam welds and the girth seam weld joining the intermediate and lower shells. All core .
region (beltline) welds were fabricated using Weld Wire i Heat Number 83653, Linde 0091 Flux, Lot Number ! 3536. . l 9 B-15 j
l l Table B-5 i l Calculation of Chemistry Factors Using Surveillance Capsule Data l l Material Capsule Fluence FF ARTw/M FF* ART.w7 FF2 (n/cm2 ,Fal.0 MeV) ( F) ( F) l l Inter. Shell Plate U 3.437 x 10 8 ") 0.706 15 10.585 0.498 B8805-3 (Longitudinal) Y 1.242 x 10 ' 2 l.060 40 42.4 1.124 3.437 x 10 ") t8 Inter. Sbell Plate U 0.706 0 0 0.498 B8805-3 l Y 1.242 x 10 1.060 20 21.2 1.124 (Transverse) Sum: 74.185 3.244 Chemistry Factor = 74.185 + 3.244 = 22.9 3.437 x 10 ") 28 Weld Metal U 0.706 15 10.585 0.498 Y 1.242 x 10 l.060 0 0 1.124 Sum: 10.585 1.622 Chemistry Factor = 10.585 + 1.622 = 6.5 (a) Original fluence value has been revised to be based on actual cycle burn up instead of predicted (original value = 3.41 x 10'8 from capsule U report, WCAP-12256). (b) The weld metal ART.m7 values contain no adjustment ratio per Regulatory Guide 1.99, Rev. 2 Position 2.1 since the surveillance weld is identical to that used in the core region of the longitudinal seams and the ginh seam weld joining the intermediate and lower shells. t B-16
Table B-6 Summary of Adjusted Reference Temperatures (ART's) at 1/4-t and 3/4-t Locations for 16 EFPY Component 16 EFPY AR'f" 1/4-t ('F) 3/4-t ( F) Intermediate Shell Plate B88051 80.7 64.1 Intermediate Shell Phte B8805-2 100.7*) 84. l*) Intermediate Shell Plate B8805-3 97.5 76.4 Inter. Shell Plate B8805-3 Using S/C Data 67.1") 57.6'" Lower Shell Plate B8606-1 77.6 59.6 Lower Shell Plate B8606-2 81.9 62.5 Lower Shell Plate B8606-3 80.8 60.6 Cire. Weld 101-171 -21.7 -39.9 Cire. Weld Using S/C Data -68.6'" -72.2") - Long. Weld 101-124A -31.8 -48.5 Long. Weld 101-124B -30.0 -47.0 Long. Weld 101-124C -30.0 -47.0 Long. Weld 101-142A -30.0 -47.0 Long. Weld 101-142B -31.8 -48.5 Long. Weld 101-142C -30.0 -47.0 2 (a) Based on surface fluence of 1.088 x 10* (n/cm E>l.0 MeV). (b) These ART values are used to generate heatup and cooldown curves. (c) Numbers were calculated using a chemistry factor based on surveillance capsule data. B-17
Table B-7 Calculation of Adjusted Reference Temperatures at 16 EFPY for the Limiting Vogtle Unit 1 Reactor Vessel Material - Intermediate Shell Plate B8805-2 o Parameter ART Values Operating Time 16 EFPY
- Material B8805-2 B8805-2 Iacation 1/4-t 3/4-t Chemistry Factor, CF (*F) 53.1 53.1 Fluence /(10t ' n/cm2 , E>1.0 MeV), f (*) 0.6485 0.2303 Fluence Factor, FF'M 0.879 0.604 ARTym = CF x FF (*F) 46.653 32.056 Initial RTym, I (*F) 20 20 Margin, M (*F)
- 34 32.056 ART = I + (CFxFF) + M (*F) 100.7 84.1 per Regulatory Guide 1,99, Revision 2(80 l'
, (a) Fluence, f, is based upon f-r (10* n/cm2 . Fal.0 MeV) = 1.088 at 16 EFPY, The Vogtle Unit I reactor vessel wall thickness is 8.625 inches at the beltline region.
(b) Fluence Factor per equation (4) is defined by FF = f* 8 ' " *8 0 (c) Margin is calculated as, M = 2 f(o,2, g,2)l The standard deviation for the initial RTm margin term, c3, is 0*F since the initial RTm is a measured value. The standard deviation for ARTm term, o,, is 17*F for the plate, except that o, need not exceed 0.5 times the mean value of ARTm. o, is 8.5'F for the plate (half the value) when surveillance data is used. 1 i l i l l B-18
. m _ .. . . . . _ _ . _ . _.m...._ __ ._ _ _ . . _ _ _ . . _ - _ _ _ . _ . - . _ _ _ _ _ .- . _ _. .-
[ Table B-8 Vogtle Unit 1 Ileatup and Cooldown Data at 16 EFPY with Margins of 10*F and 60 psig for Instrumentation Errors ; i . Includes 1) 10% increase in pressure for ternperatures less than 200 F per ASME code case N-514. 2) vessel flange j requirements of 150 F and 561 psig per 10CFR50. and 3) pressure adjustment of 74 psig to account for pressure ! difference between the wide-range pressure transmitter and the limiting beltline region of the teactor vessel. ; Criticality Criticality [ Steady State 20 G) 40 CD 60 CD 100 CD 60IlU Limit 100IIU. 1.imit flydrostatic Irat Test T P T P T P T P T P T P T P T P T P T P 70 492.57 70 44933 70 405.48 70 360 87 70 269.82 70 460 15 246 0.00 70 424.95 246 0.00 225 2000 75 50230 75 459A6 75 416.16 75 372.27 75 282.71 75 460.15 246 411.59 75 424.95 246 379.59 246 2485 80 512.62 80 470.45 80 42737 80 38437 80 296.62 80 460.15 246 411.59 80 424.95 246 379.59 85 523.87 85 48231 85 44032 85 397.88 85 311.76 85 400.15 246 411.59 85 .424.95 246 379.59 90 535.96 00 495.06 90 453.81 90 412.12 '90 328.02 90 460.15 246 411.59 90 424.95 246 379.59 95 543.10 95 508.82 95 468.28 95 427.63 95 345.74 95 460.15 - 246 411.59 95 424.95 246 379.59 100 543.10 100 523.47 100 483.98 100 44434 100 361.79 100 461.72 246 413.02 100 424.95 246 379.59 105 543.10 105 539AO 105 500.93 105 462.29 105 385.51 105 465.69 246 416.63 105 424.95 246 379.59
!!0 543.10 110 543.10 110 519.02 110 481.76 110 407.88 110 472.04 246 422A0 - 110 424.95 24 379.59 115 543.10 115 543.10 115 538.70 115 502.79 115 431.99 115 48033 246 429.94' i15 - 424.95 246 379.59 120 543.10 120 543.10 120 543.10 120 525.29 120 458.10 120 490.67 246 43934 120 426.75 246 381.23 125 543.10 125 543.10 125 543.10 125 543.10 125 486.23 125 502.80 246 45036 125 43031 246 384.46
> 130 543.10 130 543.10 130 543.10 130 543.10 130 516.56 130 516.66 246 462.96 130 435.65 246 38932 135 543.10 135 543.10 135 543.10 135 543.10 135 543.10 135 532.41 246 477.28 135 442.68 246 39531
- 140 543.10 140 543.10 140 543.10 140 543.10 140 543.10 140 543.10 246 487.00 140 451.44 246 403.67 Y I45 543.10 145 543.10 145 543.10 145 543.10 145 543.10' 145 543.10 246 487.00 145 461.76- 246 .413.05
*-* 150 543.10 150 543.10 150 543.10 150 543.10 150 543.10 150 543.10 246 487.00 150 473.93 246 424.12 !
D 150 774.67 150 749.28 150 72538 150 702.99 150 668.28 150 590.60 246 ~ 530.18 155 487.83 246 43635 155 805A6 155 782.19 155 760.80 155 741.02 155 70831 155 613.84 246 55131 160 503.52 246 451.02 t 160 838.45 160 817.81 160 798.82 160 781.97 160 756.53 160 639.00 246 574.18 165 520.89 246 466.81 d,. 165 874.18 165 855.93 165 83936 165 82630 165 SGF.15 165 666.42 246 599.11 170 54032 246 484.47 t 170 91238 170 8 %.85 170 883.99 170 873.83 170 86332 170 695.99 246 625.99 175 56130 246 503.91 175 95337 175 940.88 175 931.41 175 924.99 175 923.59 175 727.87 246 654.97 - 180 - 585.03 246 525.12 180 997.40 180 98EAi 180 982Ai 180 979.99 180 088.01 I80 762.53 246 686.48 - t85 610.65 246 548.41 185 1044.72 185 103934 - 185 1037.29 185 1039.18 I85 799.65 246 720.23 190 638A2 246 573.65 190 109537 190 1094.07 190 839.63 246 756.57 195 66832 246 601.20 - 195 1150.22 105 882.84 246 795.85 200 701A6 246 630 96 200 1208.94 200 929.17 246 837.97 200 630.96 246 630.96 i 200 109231 200 837.97 246 837.97 205 663.11' 246 663.11 205 1149.67 205 883.21 246- 883.21 210 698.05 250 698.05 210 1211.10 210 931.85 250 931.85 215 735.63 255 735.63 225 1276.80 215 984.06 255 984.06 220 776.14 . 260 776.14 220 1347.74 220 1039 90 260 1039.90 225 819.75 265- 819.75 ' 225 1423 Al 225 1100.16 265 1100.16 230 86631 270 86631 230 1501.91 230 116430 270 1164.70 235 917.23 275 . 917.23 ; 235 1591.99 235 1233.69 275 -1233.69 240 971.48 280 971A8 240 168530 240 1308.02 280 1308.02 245 102939 285 1029.79 245 1785.26 245 1387A3 285 1387A3 250 1092.22 290 1092.22 , 250 1892.15 250 14723 1 290 147231 _255 Il59A7 295 1159.47 [ 255 2006.55 255 '1563.28 295 1563.28 ~ 260 1231.23 300 1231.28 260 2128.76 260 '1660.77 300 166037 265 1308.60 305 1308.60 265 2259.41 265 1765.08 305 1765.08 270 139131 310 139131 270 2398.54 270 1876.53 - 3to 1876.53 275 1479.80 315 1479.80 275 1995.33 3 15 199533 280 1574.52 320 1574.52 280 2122.41 320 2122Al 285 1675.74 325 1675.74 285 2257.67- 325 2257.67- 290'1783.98 330 1783.98 i 290 240130 330 2401.70 295 1899.51 335 1899.51 300 2023D6 340 2023.06 [ 305 2154.59 345 2154.59 310 2294.65 - 350 2294.65
# % k W 4 3
_ _ . . m _ m_ ___
1 MATERIAL PROPERTY BASIS ) l LIMITING MATERIAL INTERMEDIATE SHELL PLATE B8805-2 ; l, LIMITING ART AT 16 EFPY: 1/4-t, 100.TF 3/4-t, 84.l*F 2,500 ,,4i ,ii,i,,ii,,,i ,, si , i i ; il
- ' ~ ' ~
I 1i! ! i i i ii !i LEAK TEST LIMIT - f i i ! i i i-i l !l ! !Nl 11 1: t ! I ii o oen 6 i i i i ii iiii i i ii i l l' ill 1 li ' i i ii 4,4 U v ! ! i ! ! i ! ? ! t i i! !/t ! l 1. Il !/ 1 t i! ! !! I i i ! ! i i !/ i !!/ I' 1 ll 3 i ' i I i i i , i ! Ii i f I II !iI i l 11 I i i i i i i i e i i i i I /T !lI f I1 i i i i iie i
,nnn
, i , i i i i , i i i i i i It i i ili ! li ! I i 4 m vvv i e I i ! iit ! l ! il t il il l tf ! ! i 1
@ i i i i i i i i i i i ! >1 l il / ! il! i i
. ! ! i i i !tii i i I ! il I i il 4 i i i en i ! I i ! IIi'l / I li !* 1 l 6. 1,750 : l; '
UNACCEFABLE i l/' ll ll ' / i l; '
; ll 1 -- i ,1 OPERATION j I t/i 'i / i i i i:
i i i i i ! i ! t / li) !l i i ! !!
@ !ii i !I i i ii !! /t I il iI I i t i ! . ' !I L,,, 4 i ii 6 i / i . I t/ i li i i i ! I i i I e ii
--} 1,500 , , ,, HEATUP RATE N . :/ I il !
i i i ii I (/) ll'I UP TO 60 F/HR N / // /, [ l ACCEPTABLE 'l ' ' (/) i - i i ! !, !! i r I li : . OPERATION 2 1,250 ll, ; llll l; l l l lq lll , l , , i , , Q l i l HEATUP RATE ii'.UP TO 100 *F/HR N {
/
/ / / _ !/
> / i/
lI l ll' iil i i a g q ugng ,vvv ii i tw i! i/i l i i . i i i i t ! ! ii i !/ i/ / 1/ t i i i ! i i i i i b i I i i ; / t /l ; i/ i ! i i i i i i ) ! i i H !i i / / fi / i i i i l '! l
- i . '
(Q ii i . A /! ! . /t i6 i i i!! i i l g
, _ . ii.
ii i i
! /i
! vi i /
/I l /i
/i i i i i i i i i l iii
- t i
. i
! ! i
!I i i i/ ! I/'/
!> i i i i i i I I l /I !/ i !i i i ! i i C 4 -i vi iii i i ! ti 500 ,
p
- i
/ ii l 6 i i
, , , 8 a i i
, , , , i .
i i 6 i ii 4, 1 4 !! '! ! i ! . i i i ! i !i lii lii !! j
- lll{
l l } CRITICAUTY LIMIT BASED ON i INSERVICE HYDROSTATIC TEST h ,- q-- i i i 250 ! ' TEMPERATURE (246* F) FOR THE LL-L f l!l
+ii SERVICE PERIOD UP TO 16 EFPY h-
! I I e 14 I iIIi iiiI ! 1 IIIi!
i i 0 l 3 0 50 100 150 200 250 300 350 400 450 500 l
; Indicated Temperature (Deg. F) l i
\ Figure B-1 Vogtle Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates up to 100*F/hr)
; Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for 1
Instrumentation Errors, and Margin of 74 psig for Pressure Difference Between Pressure , Instrumentation and Reactor Vessel Beltline Region) Including 10% Increase in Pressure '
- for Temperatures less than 200 F per ASME Code Case N-514 B-20
)
MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8805-2 LIMITING ART AT 16 EFPY: 1/4-t, 100.7'F , 3/4-t, 84.1'F 2,500 i, , ,, i , i! , ,, i 4 i i ,i ii i i l A ! li 2,250 ; f l
' ~
i r ! i
, t i I !
2,000 i l' l i
. c_.) .
1 4
$ 1'750 UNACCEPTABLE j ! - OPERATION r , , ,
i i i / I
@ i i/ i
$ 1,500 ,
,l , l l '
l , , m ' '
;:! /! i m i ii
'i i
ACCEPTABLE i 0 L- 1'250
' ' ' ' i
/
O-i l l ' l lj
! i/ /
/ !
OPERATION t+, q_t !
.! !/ 1!!
V 1,000 l,,
; ,/l ' ' '
N o , , , i' ig,i , i i i lI i
@ \ i Mi !
O ' I?,~ ' 750 ,, i , , l i! fg COOLDOWN RATES *F/HR. ll
/ i l ,l
! i l
i l i Tr ii Wff f , i i i; ; i 500 i
, o v,u . , , , , , ,
~
20 !</ v i i ! 6 i ___ 40 M i / i i ! ! ! 60 F K ll l l i !, l l i ; 250 100 n , , , ,, i , , ii , ,i ,
!!! i ii ; i ii i!i!
8 7 !j I i I i !i
!!ji i i i ! ! ! !
0 O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F) Figure B 2 Vogtle Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates up to / 100*F/hr) Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instnimentation Errors, and Margin of 74 psig for Pressure Difference Between Pressure . Instrumentation and Reactor Vessel Beltline Region) Including 10% Increase in Pressure for Temperatures Less than 200 F per ASME Code Case N-514 B-21 l
B-6. REFERENCES l g [Bl] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988. i en D " Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in [B2] Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981. l [B3] ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Appendices, " Rules for Constmetion of Nuclear Power Plant Components, Appendix G. Protection Against Nonductile Failure", pp. 558-563,1986 Edition, American Society of Mechanical Engineers, New York, 1986. [B4] Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No. IN, May 27,1983. [B5] WCAP-7924 A, " Basis for Heatup and Cooldown Limit Curves", W. S. Heelton, et al., April 1975. [B6] WCAP 110ll, " Georgia Power Company Alvin W. Vogtle Unit No.1 Reactor Vessel Radiation Surveillance Program", L. R. Singer, February 1986. a [B7] WCAP-12256, " Analysis of Capsule U from the Georgia Power Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Prcgram", S. E. Yanichko, et al., May 1989.
~
[B8) Combustion Engineering Inc., Metallurgical Research and Development Materials Certification Report, Contract No. 8971, Job No. 708124-001, Code No. B-88051, Heat No. C06131, dated October 5,1972 and Lukens Steel Company Test Certificate, Mill Order No. 10554-1, dated April 25,1972. [B9] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from W.A. House, Lab No. P18955 Inter. Shell Code B8805-1, dated March 22,1979. [B10] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification Report, Contract No. 8971, Job No. 708124-003, Code No. B 8805 2, Heat No. C0613-2, dated October 5,1972 and Lukens Steel Company Test Certificate, Mill Order No. 10554-1, dated May 9,1972. [B11] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Hager from W.A. House, Lab No. P185%, Inter. Shell Code B8805-2, dated March 22,1979. [B12] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification Report, Contract No. 8971, Job No. 708124-005, Code No. B-8805-3, Heat No. C0623-1, dated October 5,1972 and Lukens Steel Company Test Certificate, Mill Order No. 10554-1, dated May 16, 1972. t B 22
)
l l ! [B13] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from l W.A. House, Lab No. P18957, Inter. Shell Code B8805-3, dated March 22,1979. [B14] Analytical Request # 15177, " Alloy Analysis Irradiated Iow Alloy Steel, Georgia Power ' Company Vogtle Unit 1 Nuclear Plant", Lawrence Kardos, 11/10/93.
.A
[B15} Combustion Engineering Inc., Metallurgical Research and Development Materials Certification
- Report, Revision 1, Contract No. 8971, Job No. 708142-007, Code No. B-8606-1 Heat No.
C2146-1, dated March 29,1974 and Lukens Steel Company Test Certificate, Mill Order No. l 12517-2, dated March 23,1973. [316] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from W.A. House, Lab No. P15703, Code B8606-1, dated October 30,1978. i [B17] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification
% ort, Revision 1, Contract No. 8971, Job No. 708142-013, Code No. B-8606-2, Heat No.
F%2, dated March 29,1974 and Lukens Steel Company Test Cenificate, Mill Order No. 12517-2, dated March 23,1973. [B18] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from W.A. House, Lab No. P13986, Code B8606-2, dated October 30,1978. [B19] Combustion Engineering Inc., Metallurgical Research and Development Materials Cenification Report, Revision 1, Contract No. 8971, Job No. 708142-011, Code No. B-8606-3, Heat No. C2085-2, dated March 29,1974 and Lukens Steel Company Test Certificate, Mill Order No. 12517-1, dated March 30,1973. > [B20] Combustion Engineering Power Systems In'eroffice Correspondence to A.B. Harper from W.A. House, Lab No. P15704, Code B8606-3, dated October 30,1978. < [B21] Combustion Engineering Power Systems Welding Material Qualification To Requirements of ASMB Section III, Job No. D32255, Project Number 960009, dated November 6,1972. [B22] Combustion Engineering Power Systems Interoffice Correspondence from P. C. Kiefer,
~
Qualification Code Gl.43, Job No. D32255, dated November 2,1972. [B23] ASME Boiler and Pressure Vessel Code Case N-514, Section XI, Division 1, Low Temperature Overpressure Protection", Approval date: February 12,1992. [B24] Nuclear Safety Advisory Letter, NSAL-93-005A, " Cold Overpressure Mitigation System (COMS) Nonconservatism",3-10-93. l D B-23}}