ML20080S512
| ML20080S512 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/28/1995 |
| From: | Peter P, Williams J, Wrights G WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20080S520 | List: |
| References | |
| WCAP-14196, NUDOCS 9503100427 | |
| Download: ML20080S512 (141) | |
Text
{{#Wiki_filter:J n !n l,, I 4 - y g. e ... g jf; 93 fYN.u. w ,, s .~.nnn -;e'ina.hi,iauanka h ui4W;Eddas hna h uMua j v. l ~ 1 9 .s i - 1 = l_ f,. i n ') sy. ..a '1 - i pv~ E Nk ;. I.. i 0' 4 A. , ' Ip'p: ./- 9r _'. f,.f) _ f :!' t Ly Q,;.., l: ff j] , y_M <y...
- y y:l:;,;gpp '
- nwr
.y u Y$ k. N$ .w) k. .aw,. mr;; 4,'y j6.Gn:y %4 ggld'fJ:fc MQGa&bbg:- mL ; 2 +,Y l, EdG R4 9503100427 950308 0' ) __m g,. @ ' PDR.' A D O ClK. 0 5 0 0 0 3'4 8' i kiy,. p: P PDR q
- g,
.j gj 1
s '<4.,..-..,, i. 't. : ...,.....,.,,.,. J. ' "[: 7.. 7, [ *,,. ;. O C...,. 3.... '; U ' i' G .i '.'j;...,d.-,,.,,.',7,-.g, l ;. Q ,#f. l, l.
- *'l... - ' '
r +},, ' . ( l, "' '
- i,..
,,P ;g..j...- ..... ;. ; g, '.-l '............ . j.f y, mp,;(l
- 1.,, :., ;,'.... j,, t :.
-...,,.,',v,3,,.y,.,,;'..- .j..,, ,. - ', ',. ' N 'p ',' 5...,, .e -l_ .....:-{:.<. "..,.,,... - ..l ( r ,...-:-., -. ;. - ; "y,, f g:,,.
- .e,
., u, t..,_; ;',, g..;:.: - ..p'
- ,1
>g.. u, L'. .u:. '.:. :, _v ,... y f. : '.., > ~.. j,.? , x 3
- .r t..
..,7,.r,.. ', ;, i * - t 'g..,. - < , p,,.:; ,,'.,e. <., ..,,,(t:,.,.:S. ;,.,*. _, g. >,,j. ..'.7,.,,.,,,. c3. !,. s 9w ^ + '- -
- o..;9
',[, f; (,[,... .. [, ',;,, N '. e, .e4. l. e... s.- .>p . ' ' 'k '},.,,,., : s.5) #,,', ' ;i.*;h/.. / ' '*. 's
- d. g:,,
...l. a., T.,.*:,ff,..%'.;,'..- .,M,'.: ' ~j (.2.' N h. r~,[,.[.,'.. .[,I '.h . J. f, i,k. ) * ,' h'k
- t. y r.
., ' ',..,. u.,. :* y. ' (y- , j., (,/ j;/,... a,..<1 . u : ',...: - 7.- -. ;....:' ".....-.':',s;,g,:.,'.s.c.',, y ..,.,,.....I.,,:. a."., y,,,,.. y, ,.., c. ;;,,. :%, ;.,,.,,; 4 .;, s,,4,,a, - '-. ' ' ' --jie.
- ,- +
c .. g ,..,...y. - , g ; ;.lj.,%,.. ' 3..;l,n .,.. _ ' t , ;. l g.., 7,y. .,f. 7 ' y.. :. 9 7 ] y:;;*/ *.. ..s,,.-. ~r:..... ..,s .~ . 7,;, * ' .:_,,..f..,,- '( ;,' ,, y, {-. :,., ,e ,f.. a.:. , rcs, ;. ;., - J- ..> ~' .e. n'., c. !.x...., ;. : ' y_,:1 'i l;.,., {,s
- ,,.;,.,;,,., ' Q,;.(*l:,',(,,.. i,',. j. ' ' Q, ft..,';;. 7. c ;y. r
'n
- j J
- i. '. 4: i>^,.'. < j
> q., f< :;!'..g,* _.3 ' !\\ .f t,. ', - .f'.
- e
-f . f.y, j ~. : . _ * -. y; ;;..., A.;}; ,1 +q..~, ..,), .a. q.._.... 3,,,,-. i...'.- t ~ r...-!.- , 3. >.. :.. .,.,s. .- -,,,-p. _ g.&- s,.s .,w,,, ..,. p ;,..,.-..,9: n,..,..- '.J,.... g 7 ,3
- . -.'%;."';,.,3...,,
,s, .4s., m.., c - .,,...,~ * .,c + s ',..L..:
- e. :.r q t
- .. ;s -'. ' i-:.. ";:
.,n..s.c. ;: ; ..t,..,~.,. .e '.r- .- :.,\\,. - ..', ; V. p, n, '..l.,. ;s i r,. .,...'r'. '.n*.. -,e, d..~, '.~, " * ';.Y;. r l..A...., .t,=m I,;,
- .p
- ',.,. :c;,.. V.,,g*, ;,., :;' y ; :..,, o's .s l: s,.._"... ,,,,'f.,.;* ~. j *.' [ 3 - '.; - ' ',a,..- '..:f';,s '. \\; f . R.:.4, : :... ; 7. cs:.v.,. : ,..y ,',,,"u'p;: -.,.c. .s;: :, .'s . : :;.T,... + c, y '. - .,1: 4,
- 9. -
.t_ :[ n, * - cr --
- Lm p
./ ..,...,. ~, n.,' w:, s. ,.,.-. ~,,. -(..;- . t .".;C..: ;..j e,,,,,t ',.,;:r.*+g....,s...:i .,': i, ;.,..'" '.. ' y :) 4 :.. ;. 9; ?,,,. f..?..., s..:- ' *1 ork,.i..._., - :+.. r. .r.e : ' ~,
- 4 3
',r,.n *. + :;L y. i-b.h,. F
- ,:s,.
.,, -> i s.: .,.,'.,.=,..;'...,s.v. _ a., i n'T.^,, s. ,.- o -. g * ** :':.t. :......'.. ;- ' :}: ;: {!.. J. ,'.t,.,', ~ ', % '., :'. n,.,. 4,,':ll o .n.; \\ ~,,lm. vo.
- __,-,'_,?. y;.
- . %,, r y. c ;-.. d ). J ',, _'. c
,1- , u ' 9 f.,*, v.. \\. : >
- n. '
....',,L (s.'n"f'....,,i.,...,' g; e-. .<:\\ :'.. ;e r -s .. <. 3y~..... ' -t..' . -;; s.:... - .c ,w.,...f.,,>v y ;,,.c9,,. a e,.'.,,;,... \\,._ s,. ' ' ; .s .i, ,.',(,.'n., .'d...v.- 3, ..,, 4 3, , ;........3....,....y -..: y..x, ;. : o ,... e,f,.p ;., p ,,e s 'y . p.;.. _, * :. : < g,, m ,.;,.,,,4,; Q M.:,, - *;,.,. . j ;u i. : ? 4:*
- --.,'.. ;. Ve.,
\\'.J .. 4 ;;;*._3 (....,. n.'...,:- 4 - ;:Jt, ,.; j;; .y, ,. c, g,,,,a'
- 3. ',,,.- p. p; ;
s e..,yf,f,, ~ u.',,, 4 y < q,'. * * -,, L 4. :.s .,,:r ',..,K'_.A-'.*l.P,,,,. g.. j,q.,',{,',,',,',,.4;'.' .I 'h '# ' *** ' -', y",,.,' gi- +, ' ,.g s-
- .t
.t. 7,., 3:,, g.,, .3 ..,.,.;.,,s" ,;..'g. ;; j' 7., r ,.,,; y j. ......e.,- . :, ) u. ' -.. k ;.'.. _ _ s..y,. n ; ; .. -.., g '.. ::..,. : 4.:1 : y ',C :Q*,.7. Y : '.,.., ,' I .,L.. :. ;- L./k '..; ?
- 'Ea
+ ' 4') ' '
- 4'+'*-
9 i ,5 ' ' " ' '. ' b '.$ 1 ' O. \\ ' - *'
- O
\\ *k '8W .I- .r= ~ ^ '-'*I L. '..,;,- ' ',, s. _ q p.:. ': q,,'bpG,,;,,.];;y:.,% 1: y ; V::.. > ' H:;,..,,,99.. :. r : ^ g,,D.:
- .f.
S -(7,,.,. f.... ,.s .,j,,,'. .,e..,..,',.':. n.._ ' ;,:,:; y -l c ,,.g,- .J.,w,,'.,,,',4.j..l,,g,.' '[ .... ; ? ', 3, * /;, ,i,. y. e i,
- :1 :
,d', a- .< :e ' b:" ;, ' .*b'.; ':.;;... : s.. : . f '.,1 'I
- s......., -
y
- l. ': R.,A
- f;. i),{...*.
't $ ' gl[ '...' .}, ' ; -._',;,, ,, - j, _g
- =,y, 9.,..
e..;.. ),, ,,, v: ' ..,. -+g ..-m; ..',',g...",_. - i - ? ~;.,.-'
- %.p.
..,.,, qvT..,,.*,zh. ', y a'. ,..,R ; t',":, }.'.',.....
- ....~ :
= _..3 'a, s.- ,*. '.* ~ \\* '* ' r ~: d ,.[j',l'"_. -? (. )', r!. .[; l '%+,je,1 [.. 8'
- r, h,%
- ., '_,,,.., l, O
.; ',' q '...,.-l.....*...I a
- y,
4,$ 1,.s..,.o' + [
- e,.
-..'I'*s 2,..'<., e# ', i J. '
- _.4
$.y,j.,l )) '.',9e,* ' _ lF,.* s',*, ,g '... .q y - ~.
- l
- * ' ', * *,~~,. :,., *.,,...-
'y';.,.:*, ,c,.. f. -R. .'. ;,* .; -(,,)
- ,.'..'.}*., _. ** l e.,.._
g ,,,;-,~,.,-,. ,h;.., -. g;...... -, .z :,, .g'. l 1.,t,, q t:. 4,v
- 4. s.,'z.-
,..e, n .f ,,,f
- _ p,q
. u- . x. ./ ,.,g. .,t .: ;.,. s: ;., _....... r a'.
- y..
r. ,z_ '..,6 a ,_,. * ~ ~.,.*. os....
- m...
? f.,.,.. e,. , - ; -.'.?,. . ~ + ; 1 ';,..f- '- :,..,. ,n .,,'./,; -t:. ,..,3,.**
- ..n. _" *'.*.-;. 2 :*,-
s' a' I* w.' :"'*- 'Vi -c. u.'..... ' ..c...,...,.;., .,i: - ',. <*4 :; * :..,. ' p.. <..:'.',t.....d.'.'s r..,.a. ,4 r ,,x,, - 1
- 't'.-
C, :. n.. ;, '..' r ( 5W.;. ::t..,". -. y q n'., :y.J., * ';.~9.' 7 ,' ?..' (; %,.L i.: , :' 1 %. ,...&,:,,..',.q.".'Le,, ,,.,i,/.... s.,e ,.1 ..,o..y. a....... ,%.,:,,.. ' *3,,re, ..'..-.a .,',..'.,s c, '.~..;. < <:g e.. z. ..gce.* ,f.','. v. s m ,- - s...t;. T
- .,;s, 2,).. c., t..y,--
'r 1 s~.. s ,v*.. A ;.,. ' ;, . f _ f* N '. , ....v' <. ~.. C 1 1 .t.*'- ?;. '. ' '.,. L.- l.; f, '~ ' - -,. r,* g' M, ), ' f.y:., ',. ',. f'y;. _'.':,,.. 't -', _' .s*.,. 4 s G. L..,i J.3,. '. c A f: , ' ; j ' U,;.,:..l l ;,, ',; ' :;y.' ". 's.g. l '....,[i.;j:# q 's y.. ' y-a..- . ;._ ' f,'s., >;..,:; *. ::',. j..).[.,..- -'. f. .A*e +J , :,. w :,,.~.;;...':+'.,, :.' ' ",. : ;./. =.Q(.I.I3..,f.+.'.. + VJ. M'., '*[' '.,., (,., ' *.. L, ,[J ',.. .'i 3;} -' .o .y ..: 3 f1, (' '. t..; ;'. :,., as ?,, :. >. .eW'.;_ ';.'.. J,..,. ? - '? '.:y.. .,4 ...,,.;..;.. i. ;.., '. Y : .r.,.,...- ,. ; ;, y,,:,y,, ',,;0- ., ".,..l..:% .,,;..L'. .l. i ~~; ..: p,. .1 .q.J,.,, i ;,.,f',..,...,._.,,u. .c._.v. -,.3.... 4, ~, ' v. } lr v ..- r m ......:..~:. .y.. + * :' ;.s.....- - * ;,'.. [,.,...,.,,.- . '....r. 1 n.'...
- , a
.o ..~. *.';.. '.... -.,,. a . _,e I .-..,.,c..m.....,.. ;;s,._..;
- ^.Q....?...
,* '.]','.$',',-;'.+. '., '.. '.,....:.* ['. ?,.
- '-l.,^,.,*-.
4,, t . '..[,. ..,,.;s'., s. .,,"].~ 4 I, ' i;.[ * '., "' ;* 4, .s .f.' v,..+ * ;...., '., ',...'.'*. .. '..s:s:: .e. 'j I..,.....'., j. ',:. ',.
- e'. Af:
,.i.', 8 .O.....~;...; r:.'; ,r. ':..:, } m; .,g i
- . l
~..,.. Q ; 6.g Q',. .d ..,;,,.;,,.q' .M .,.s :: '_:,.,, . -. _. '. +i ..[(4, 'l,. 6 :s.',, _.. '. o ;W -' - ....a ,,1 l,.. .';- -* r. u. ~'!. ,_s... ,.q,,c' '. ;.. 3* -,.,'.y... d' .r '. O.y;:. ; o s ..'y* ,,,';..,(<,_ '.e,,,.. . a ', y j l . -) c, '. n:L.L :
- s
,v' _ ~,. -p v, es ,., -,,,%,p::1 . _., j' :i ~;a.:.. i.,.;,5,. ; <. ",. p y;.);y. ~/;Q', ;.;; ; y, t;;,.,, .;g.. i :,;.., ';_k, <,, :;. K..,.,( , y;,,,' ...v^ . T h .- ~, . '..,. y, ; .- } . ~:
- ..,. ; e... <..f.G..:.,. ;.'. ; >:
.f;,s.,.f'..,.,,.;.,. <W:. _ t.. :.,, 4,7,,'; 9. <:p... +.. :. - Mf..;;4,m,y ....,.;.: -Js.Y. t. p,.. .y-, 1 u f..i(. g,.. ' '., .t. .e .,.,t. .,..,a.-
- +.< ; f. ;
,g. c. -. _ i.s .,p......i., 7,., < - a,oe .; y 3i ...;,,s-.. , p, ; i,. < .:,.^~,....*3 ,.s - sp o -),.,,.,,...6, p,.,. .,.g...,,,.,., i .J e, v 4,,, ..,v.",,.,-,.,,. c.; t y, =~;., q~.. -a.s
- .s.,,..,,.
p'. ' 6....,' k,.,3.,. e,. y t.. 'N,.,,...,+p, .,e E.',.,s...,.., 3 1,: , i .- -^ :.,.;,\\l...',a.,' . ; ::., p,.. .. u, ' l..j a..'.... ,.,s~,,.,. 1
- .,.. '.+, u.....'.
- .,....,,..'.,;..,'..*.y.
- .. ; ;,,,". ; m
. ) +. : 'l$. 1 b;. ;, - 4....-'. ,.; y,.,e,,. . :( ,e,,.,.... s..,,,. v :,;g,...,,...-'* ,,,u s ;.,. :. ; b r. ,. ;,;, ;, s,.: .g c 3.. .5.- ]. ,j../,,_.',.;*(,._. .g ,,...n . ;,. ' $.,..;, ;fc .,,........t .2 , '.g.,.,;.,.. y77,:, ' y ?.,f.".,, p'; ', r 6 f,,e, g ); s. J' ',,. a-y fs.:. -ms J..,._, t.,.",., /. es g* - n, 4...,.c. 3 ..a. , A. f. ',, 3 ,,g... j.- .!.,,-.,,.2. i _ ,....:._.~..,,<,',,,y ,.v..-.... .. ' ' -- l4. ' u;.. 6 .,c.,~.,., .,4,,,..,..,3,.. -. - - :' - a'
- - ;, y s..,.x.5,.,,,
" ' C. ).. e...,i.. -,.,, ,,,6,a,: ..t"... L.,a.a,,.,:,.m,. O, r. t. 1 - < 4- .. - c, a s,; f c....sq q,;..'- e
- ,.,l.:,, ; A's..:.
><.g.A -..;.. - ...w.. U,. - 1
- a...
.+ -; y s. f<, - z
- .....S..-
p s;.' y
- l;,..
,,,.,.~:.,..-..., -.,. q ' v.,.], 2,.. .. ; ::. r.,. r, '.. i '. ' c. '4., '1 ,,..' ; 4...+.. ' ^e' / : ). A <l0. '. -.4 '..i.'-~ i .? .c u -. *. i,....... j g. f. ;... ; w. V s.' .:.2
- -' R. *. '... - :
d, J.s,.... -.,'.. ' '. .4 ; {; (
- . /., .
. '. '.,*r. +.,..t !.,..~;.:..,.,'3.,- f,.. .....L.o. '.f' ', ~W. *..:.: ~ 1.'. y :. ...' s...e... .mg..
- s.,..
..:c.,. ; .e.1 .V y ,c .'a ,; %.. ';W{ * *c ...., '.c. *. L. 'l<.~.- g i- .1.., - c-
- ' 3' '
- l.
., ~
- i; * '. '.
e .s.* .c
- .~
- 4. L * -.G :,, ; '. 'g" ;..*
-,e,
- *.... * *..,.l.
.; i,..
- b
- .,., 's'
'O .e L'.-'.*~,.C ' G ' *: ' .. g:
- 2. -...b; l.
.L,s
- .,2,,.~.m. *m.f.'.'_.I.'c
.' : 2'. ' O 0,.,;. :3 t;.' ' N el> l '.'l.: * ' " t o J..,;, ' _., -,2 '""R,. ?.. la ~,~.:"'
- c
'c~ 'q,.. ~, ' l ' '* :' ' *= :':^:n: ^i,',~~'4
- e.: s', ' ' ' :',L., g' : _ *, '.
-. ~ V {. ;,,. I. ,~,,..l,' Je* p.f h ' ',';' C +.;,'~.,;..,. e.:,..,,' g.,,..s. ...'c- . ;;o.:.s.,... g,: y'...'c.g;
- ,,,v'
- u;...*,,s, ;.,., .,7.s,. g' ". ; t.,, f.,...g,. h >.c,-... u,: %!. "..*y;.- ,s7. t. -., -w .m '*s ^ 1 % ; < * ),: a ;,....- ...(... .c.g 1 L;, [ l - +,^,:,. -.<al. _. ;, z.. h,,,, ' ' o '^' '. ', ' ' _ ;;,, -. - 4 tt e.y: i.,,,.&, .... g. - ,-i -v4 1.'.... . ~.y,. v :Q ).;, ;:u,',. , q .. T,^ ;n. ",, 2; . ;6: m.... F.. K. ' G,1::;. s' i -,. % },.'- ,M,y s -.,". y[..s ' - ' :: .'..'i*.- , h,*lQ ;: - *A ' y'.; '..
- ..: s '
y. '....'.t..' 'l.'.. y, ',; ~ 6':. -- n~.... n; . L, * * '. '. '.u'. l g ',[-,._, :
- ',,.. M,.
d ,. ~.. .*,e'.L'. i .,j '.g.., ;4 4 Y.; L .-v ',;*s_.'. ,'.'g g,, ,. b -; :. ; .,,.. i ;., ?, ',,.- } b_ , 5,.y,. q '. '. _-pe.- A,".,~.f '. _. -. y ..1;'
- .. *-,. n, ;,.
.,,4 , j :,_ '
- .l ;; - ;._,,
- ,_.s
- V*-
- 6., *r:Jf :.
., ' \\.; + '.. ;-.... .g r. .j, ( 4.,',,'.,w.,,;' s. ,.-,.~ ,*.,x 1 ,,.;.. t, ; c.
- ., c.
,q v,.L.'...~,.L.,... 'J,',.
- c..x,.-
v... .\\,.. v
- -,',",.p,
<-.- :.,;. )... -. '
- 5. y
.f.; .,s_ ...y;..,g,.,,g.,.,,...,.4'.,..p,- ,t +. 7'f.,.,. ,'s. .,, ( s,,.:- -,,. *.- - -. /.,., ;,, 1.':." t s ,i.,y.,..._.:.'. .,'. p,l,. q:.,,t..,.., 3 ;. ;..-a ...t
- ,,-...,,7...--
,;.r.7,,.._...,._.,,,.,, g ;; , ;.7. ; ,Q.'>,,,,,.,t. . ; f. y '. ' '. ; '. f. ',:,', 3._.,,,.y
- f o[,;
M. :' :3. -. _, _ c .:., _.., l ' j ; '. }. 3.,. [, : y. s?.t.' '_l
- l
_,. ~,yt'.r -.r 1 ,)., ;..., '..i ? ' _ R', ;.e *.. -..l V{f v
- g. /. :;, -.
i..b~.*....a.,A ,...n ., F._c ;7',.*.',',l>h.. a....l:,,'.... .':+> +~,:,f~..,. 4%._,,.:+'.:....,1. V,. ;, l,..,, :.:. *
- 3...:
.,',.,,.'.,..,e,..'
- 1..' ;:;:. 8 -; - ;
} ,s.. A,,'~.',q-L...
- ,c...
w ;..m. ;m, fJ..i ; p.e.,,a
- .. h.,...e.' N * &.
- y'..... '..
.,3.,:...,. y;.* ll.., .,f'..,. ;,~. a
- 3). ;.g:j
- 2..'.J,,.,;
.,,, _., ; 5;,, ..,.. j 9..,') j. -( o y * . e' 6' 'l.,. _s.. b',f ; ; b, I ' f. L : ;'., _ ' ',,h:. .'...-l::. .:. 'g. '. '., ;,. -.l. ,:. '. r ,y ,.,t .g, . 7. ' ..:<i,+.*. Y"l['. '$..7,i.%..i,
- ,:.b.
'. '.)..'.e. '. **~7 '.' h l' ':l* *ii.Y '^ ~ s. ...-v.... ,(i,...-
- ;, l M
.c 4l
- e A
l ~~ .lC;.'.\\,l'h:; '. k '..'.l.. '. :'..',$ "'.( V N Y. l :. ~ . l- ** }
- .;$ 0.;
1 % ?l 1 o A ', ', t :.',.;, lz, '.'?;. 'f-. ". 8.;, ' l. u '., %. ! . :y'. .T, r i '.;y.- y-.',u.. b..,
- s. ';Vl l ty -..
u.% f.
- .-..'[.,.,...
.',...3,.(,h,'s! . +. l ;... ! -f 8,., i.:,, . g.,..,.. - 'J.. g. q. +*1. '.4.,,4._;r*"... ...).....
- ^ *
-'..G',..'i:.:~..*.'.'..'l,..,
- 4.L
- .~ ) ' 7 '; *
- 4
.i.,.L, O.
- P* :'t 1
- +
v. J. ~,s' y,. y '..".. ", ;... :t;.s. ; q';, '..o,.;.' sl],. '. *, /;*lg '! 1 l E :), ':, *.& E,:h','.,.;. as. :. ,,,(.,,.,*..',.3','..-.3,,..%,, _ g.;.. 7 ex'( y* ,' f - c: .. A';'. s q~ - ...,... '..': s. :.,.',r. l.. * -Nl ..c. p. '. j,7 . :,,. ',z.., f, %, '.*& ' :,.. '..,., ' '. *.;.;: ',..<.~.x- '_ 'j. 3,m ~ l * :. _;.. < '
- o
,yl...,..,,".,"p ^
- " * \\ ' *, 3+:[,y
- .e
-' ') y y_... : 4-s * .,f' L,N '.7, Q,' ll ;,..O ' y ;9 ~_;.;f_ ' . l.';* ; i.. t.,.....:; * [.A.,., V. ~j;^.":. g, - A. 'S ? 'l. y _. ~,-f...!..,;.,. s :'l,: * ;'" ',: s. ' .t l.* ,.. y l .... '.p
- 3.,f, :s.,.1 r,). *, ', ', ; [',
.., '. :. : ' ; p'.;;,.. p *s S.
- f L'
- '..,*g,
' n. !... a. }. q w.*:,\\..' * :.i ^
- .; 4. ;.. ::
..f_.a :.' h,.,. s. *.,- : .., :s,.a." .;4,.,..,.' y l,,. .. y .'. (;, -, lj1 qn-;-* .r . -.-s.,:., a,,;., g. ,g. ....;-? .9.,.g,...,l.. . 'l;{,j..l,..... :. ;.y " r - :,,, .,:7.,. n y;, V.,..,,.::.; ;;,,. i s . p,...., <g; n'. 1 lf y.., >, - -: :. ;... w,,,.. " d.... ? ~ 7 .: _I. u.' a ^ *.. ,a, I*( [. 'l J.l ' : '; ;lh ',' i.' s\\- E ." 4 l, i ' ' f 'Q f. : ! : t. t..f., 2 7,,,'.'.'.<'.,c.*.:."*;., ,r...- 2,x, i,..' : ,,.,..t- ,..'"f .nq... . i '.,..*:.. '., '..1 :, .6 ~ .w ..3,a....,.,..,,. .e.--a f ~ q,;,p' .y ?. -; ' t ( f.,, j i.,;... ".. g.,p. ~ y '.'l (g' ' f.,;,,, a*.,,.'.,. ,., n. 9..,.f...,,. ; f..*.., ;. l, - * *,,,.y,.;, ;,. * ;.. p. 4 .." 3 :,.
- .,., 4.,
. ;v.,..,' . ".,,...}.. p,;,.
- y. ',4,g,,,,. r.. ',...
h.j.) h ')f ' ** q*....r.c.,. I, y,.,. &..j *, ' c,. - '. ] ,y ,.7, '.[*.3 .'4.*',.'3 .[.'.,,[.. '.i..*.. 4... ..d '.w' s
- p. '..'.$
q.' .i: :*,.'t'. 3,. ;'.". ' g ; - ;. ,,e y.'>.'.I,.,.. } } ','.l.- [.,, ', ' + ",*',;"'.,",'/ ' lj;'2*(.,J.:y 4, g, Q - g.,.. ' 1, g.' *,' ;,!,^. 5.k,. ,g., ,.1..%' 7 s - + '* ' f, 9.[g'; '.f. *r, + .'I' '. j ., ';f.
- ,f e - fl.jf'
- :,. "< ' h,.3.'.; c,. j ',.'
',. j 's [ .. r.'. ? i :. .. un.v.w w w ? m..:, ; &..~w.. 8' ..s...,.,.. , m w. u g a;i:m.g...,i,fl,4 g w,g 4,l.nl.:u x.l.d;j },&,,;'.;.T..,&., , _,',.g.i, '. {.*b,h ;j q. r,..). -','.&r::[, s al, :y..R+. p u..s.n.s.....g.c. y v.. ; :.g&x. we.,..m:m..,. :1.. w.w;..
- . m, A+.......w.,...s v.s.. ;
- r g
- 4.
c. a.,.m.. . gu. 4s. %,7,n.o. <. r. m q:.g;' :. : .4. j.. 4'a w 3 wt. c n :*. n.,.. m. q. 2 c -.n,..,.- . eJ. ',j,..(,1.. g w e E }.~,;f]f,s..,','l.,';,*' hi ;]'.;! l.;,l,".1 ?;3 q.4. lD... 2. +..vN U:j ['f.R. [ t l.fs..R; <* ;' %50$s. 1... ..g.
- M.
e., .e.1 g, ... - i .73.- rid I .&,d:', *1,i. :0*A,' [., j 'Q,'A Yl.. l .^
~ .; F ' Westinghouse Class 3 (Non-Proprietary) ? WCAP-141% f ANALYSIS OF CAPSULE W FROM THE ALABAMA POWER COMPANY - FARLEY UNIT 1 REACTOR VESSEL t>- RADIATION SURVEILLANCE PROGRAM L. P. A. Peter G. N. Wrights J. F. Williams February 1995 Work Peiformed Under Shop Order' AWEP-1Y . Ihyered by Westinghouse Electric Corporation for the Southern Nuclear Operatirg Company Approved byN R. D. Rishel, Manager-Metallurgical & NDE Analysis WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1995 Westinghouse Electric Corp. All Rights Reserved
- f I
1 l i e t i ? \\ PREFACE 4 c') H-This report has been technically reviewed and verified. -i Reviewer: Sections 1 through 5,7,8 and Appendix A E. Terek M; r-i E. P. Lippincott d.[OMb03 m fM EPL l! Section 6 - 0 l t 9 f k -t ii: ~l .k i i:I t b i ~f k I l 5
l 4 m ~ TAELE OF CONTENTS 1 i .i I L Section ' Title .P. age a -I LIST OF TABLES iii.. j i i LIST OF ILLUSTRATIONS vii ~ 1.0
SUMMARY
OF RESULTS ' 1-1 2.0 ' INTRODUCITON. 2-1 I '3.0 -BACKGROUND .3-1 i 4.0 - DESCRII' TION OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE W 5-1 5.1 Overview. 5-1 5.2 Charpy V-Notch Impact Test Results 5-3 5.3 Tension Test Results .5-5 5.4 Compact Tension Tests 5-6 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6 6.3 Neutron Dosimetry 6-6 6.4 Projections of Pressure Vessel Exposure 6-11 - 7.0 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE 7-1
8.0 REFERENCES
8-1 APPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS A-0 u ii
1 + 1 LIST OF TABLES m ^ .Titis. T.1111 Ense. 4-1 ' Heat Treatment of the Farley. Unit 1 Reactor Vessel Surveillance Materials '4-3 4-2 Chemical Composition of the Farley Unit 1 Reactor Vessel Surveillance- '4 Materials 4-3' Cu and Ni Weight Percent of the Farley Unit 1 Reactor Vessel Behline Materials ,4-5 .c 5-1 Charpy V-notch Impact Data for the Farley Unit 1 Lower Shell Plate B6919-1 5-7. (Longitudinal Orientation) Irr=ds=*M to a Fluence of 4.040 x 10" n/cm2 (E > 1.0 MeV) 5-2 Charpy V-notch Impact Data for the Parley Unit 1 Lower Shell Plate B6919-1 5-S (Transverse Orientation) Irradiated to a Fluence of 4.040 x 10" n/cm2 (E > 1.0 MeV)_ 5-3 Charpy V-notch Impact Data for the Farley Unit 1 Surveillance Weld Metal' 5-9. 2 Irradi=*M to a Fluence of 4.040 x 10" n/cm (E > 1.0 MeV) ~ 5 Charpy V-notch Impact Data for the Farley Unit 1 Heat-Affected-Zone (HAZ) 5-10 2 Metal Irradiated to a Fluence of 4.040 x 10" n/cm (E > 1.0 MeV). ,l 5-5 Instrumented Charpy Impact Test Results for the Farley Unit 1 lower Shell Plate 5 B6919-1 (Longitudinal Orientation) Irradiated to a Fluence of 4.040 x 10" n/cm2 (E > 1.0 MeV)
- - ij 5-6~
- Instrumented Charpy Impact Test Results for the Farley Unit I Lower Shell Plate ' 5-12 l B6919-1 (Transverse Orientation) Irradiated to a Fluence of 4.040 x 10" n/cm 2 (E > 1.0 MeV)- 1 iii 4
p;( = 7 f,, 1 l 3_ ?l D LIST OF TABLES (CONTINUED) x .l Tab.h Title - ENLe. ' 5-7 Instrumented Charpy Impact Test Results for the Farley Unit 1 Surveillance-5-13 i . Weld Metal Irradiated to a Fluence of 4.040 x 10 n/cm (E > 1.0 MeV). l 2
- 7
. 8 . Instrumented Charpy Impact Test Results for the Farley Unit 1 Heat-Affected-5 j Zone (HAZ) Metal Inadiated to a Fluence of 4.040 'x 10 n/cm. (E > 1.0 MeV)' ] 2 ,9 Effect of Irradiation to 4.040 x 10 n/cm (E > 1.0 MeV)'on the Notch 5-15 { 2 t ' Toughness Properties of the Farley Unit 1 Reactor Vessel Surveillance Materials j I , 5-10 Comparison of the Farley Unit 1 Surveillance Material 30 ft-lb Transition - 5-16 j Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide l 1.99, Revision 2, Predictions .. t '5 11 Tensile Properties for the Farley Unit 1 Reactor Vessel Surveillance Materials - 5-17 j 2 Irradiated to 4.040 x 10 n/cm (E > 1.0 MeV) - i 6-1 Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center. 6-14 l 1 . 6-15. j '6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates at the. Pressure Vessel Clad / base Met 1 Interface 6-3 Relative Radial Distribution of $ (E>1.0 MeV) Within the Pressure Vessel 17 l Wall j .t t 6-4 Relative Radial Distribution of $ (E>1.0 MeV) Within the Pressure Vessel 6-18 j l Wall. j 6-5. Relative Radial Distribution of dpa/see Within the Pressure Vessel Wall 6-19 i i i 6-6 Nuclear Parameters Used in the Evaluation of Neutron Sensors 6-20 l l IV j -l
LIST OF TABLES (CONTINUED) Tab.le ' Title Ease 6-7 Monthly Thermal Generation Durmg the First Five Fuel Cycles of the 6-21 Farley Unit 1 Reactor 6-8 Measured Sensor Activities and Reaction Rates, Surveillance Capsule Y, 6-23 Saturated Activities and Derived Fast Neutron Flux 6-9' Measured Sensor Activities and Reaction Rates, Surveillance Capsule U, 6-24 Saturated Activities and Derived Fast Neutron Flux 6-10 Measured Sensor Activities and Reaction Rates, Surveillance Capsule W, 6-25 Saturated Activities and Derived Fast Neutron Flux 6-11 Measured Sensor Activities and Reaction Rates, Surveillance Capsule X, 6-26 Saturated Activities and Derived Fast Neutron Flux 6-12 Summary of Neutron Dosimetry Results Surveillance Capsules Y, U, X, 6-27 and W 6-13 Comparison of Measured and FERRET Calculated Reaction Rates at the 6-28. Surveillance Capsule Center Surveillance Capsules Y, U, X, and W 6-14 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule Y 6-30 6-15 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule U 6-31 6-16 Adjusted Neutron Energy Spectrum e the Center of Surveillance Capsule X 6-32 6-17 Adjusted Neutron Energy Spectrum.A 6 Center of Surveillance Capsule W 6-33 v
1 - LIST OF TABLES (CONTINUED) F Table - Title _Pjgg r. 6-18 Comparison of Calculated and Measured Neutron Exposure Levels for Farley' [6-34 Unit 1 Surveillance Capsules Y, U, X, and W - . i 6-19 Neutron Expos 6re Projections at Key Locations on the Pressure Vessel 6-35 Clad / Base Metal Interface { l i, 6-20 Neutron Exposure Values for the Farley Unit 1 Reactor Vessel 6-36 F-l 6-21 Updated Lead Factors for Farley Unit 1 Surveillance Capsules 6-37 j 7 Surveillance Capsule Withdrawal Schedule 7-1 j i j' ' t - i ) i r e i i i I Vi 1
7._- m_ ,t.p, f LIST OF ILLUSTRATIONS h ' Title. -- h
- 4-1L Arrangement of Surveillance Capsules in the Farley Unit 1 Reactor Vessel.
4-6'. 4-2 Capsule W Diagram Showing Location of Specimens,'Ihennal Monitors, _
- 4-7 and Dosimeters b
4 5-1 Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Lower 5-18 Shell Plate B6919-1 (Longitudmal Orientation) '5-2' Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Lower 5-19 Shell Plate B6919-1 (Transverse Orientation) 5 Charpy V-Notch Impact Properties for Farley' Unit 1 Reactor Vessel .5 Surveillance Weld Metal 5-4 Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Weld 5-21. Heat-Affected-Zone (HAZ) Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Reactor Vessel 5-22' Lower Shell Plate B6919-1 (Longitudinal Onentation) l 5 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Reactor Vessel 5-23 l Lower Shell Plate B6919-1 (Transverse Orientation) f ' f 5-7 Charpy Impact Specimen Fracture Surfaces for Parley Unit 1 Reactor Vessel 5-24 l Surveillance Weld Metal l 5-8 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Reactor Vessel Weld 5-25 -) Heat-Affected-Zone (HAZ) Metal 1 i ykk ~ i 1
7 3 1 LIST OF ILLUSTRATIONS (CONTINUED) : t flgge_ Title
- Page, i
< 9 Tensile Properties for Farley Unit 1 Reactor Vessel Lower Shell Plate B6919-1 5 ! . (Longitudinal Orientation)' i . 5-10. Tensik Properties for Farley Unit 1 Reactor Vessel Lower Shell Plate B6919-1 5-27 (Transverse Orientation) 5-11. Tensile Properties for Farley Unit 1 Reactor Vessel Surveillance. Weld Metal - 5-28) ,I t '~ 5-12 Fractured Tensile Specimens from Farley. Unit 1 Reactor Vessel lower Shell 5-29 i Plate B6919-1 (Longitudinal Orientation) j t 5-13 Fractured Tensile Specimens from Farley Unit 1 Reactor Vessel Lower Shell 5 ! i Plate B6919-1 (Transverse Orientation) t 5-14 Fractured Tensile Specimens from Farley Unit 1 Reactor Vessel SurveillancA. 5-31 Weld Metal l t 5-15 Engineering Stress-Strain Curves for Plate B6919-1 Tensile Specimens AL7. 5-32 ~and AL8 (Longitudinal Orientation) i 5-16 Engineering Stress-Strain Curve for Plate B6919-1 Tensile Specimen AL9 5-33 (longitudinal Orientation) I 5-17 Engineering Stress-Strain Curves for Plate B6919-1 Tensile Specimens AT7 5 and AT8 (Transverse Orientation) 5-18 Engineering Stress-Strain Curve for Plate B6919-1 Tensile Specimen AT9 5-35 (Transverse Orientation) j i P viii i i f 1 i
xn t: p,_ s e t I b . LIST OF ILLUSTRATIONS (CONTINUED) I l BIME ' Tillt Esse
- l i.
L5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens AW7 5-36 and AW8 e 20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen AW9. 5-37 -i 6-1 ' Plan View of Dual Reactor Vessel Surveillance Capsule 6-13. t -i l l 1 I J t 'I i f ? t I l .j .3 'l i i i .. { t ) 4 i . I iX
0 L SECTION 1.0 -
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule W, the fourth capsule to be removed from the Alabama Power Company Farley Unit I reactor pressure vessel, led to the following conclusions: The capsule received an average fast neutron I aence of 4.040 x 10 n/cm (E > 1.0 MeV) 2 o after 12.43 EFPY of plant operation. o Irradiation of the reactor vessel lower shell plate B6919-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction (longitudinal 2 orientation), to 4.040 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 155*F and a 50 ft-lb transition temperature increase of 160*F. This results in an irradiated 30 ft-lb transition temperatum of 125*F and an irradiated 50 ft-lb transition temperature of 160 F. 2. \\ o Irradiation of the reactor vessel lower shell plate B6919-1 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction (transverse 2 orientation), to 4.040 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 145*F and a 50 ft-lb transition temperature increase of 135*F. This results in an irradiated 30 ft-lb transition temperature of 160 F and an irradiated 50 ft-Ib transition temperature of 205 F. Irradiation of the weld metal Charpy specimens to 4.040 x 10 n/cm (E > 1.0 MeV) 2 o resulted in a 30 ft-lb transition temperature increase of 95*F and a 50 ft-lb transition temperature increase of 90 F. This results in an irradiated 30 ft-lb transition temperature of 15 F and an irradiated 50 ft-lb transition temperature of 35 F for the weld metal. o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 2 4.040 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of t 110 F and a 50 ft-lb transition temperature increase of 120*F. This results in an irradiated 30 ft-lb transition temperature of -40*F and an irradiated 50 ft-lb transition temperature of - 5*F for the weld HAZ metal. l-1 e
a,- w k/ 'f ..[ r 1! [,,
- Irradiation of Idwer shell plate B6919-1 (l$ngitudinal orientation) to"4.040 x 10" n/cm (E S -
2 o ,n [ - 1.0 MeV) resulted in~an irradiated average upper shelf energy decrease of 31 ft-lb, resulting
- in an irradiated upper shelf energy of 109 ft-lb.-
I e 4 L o= - Irradiation of lower shell plate B6919-1 (transverse orientation) to '4.040 x 10" n/cm (E.> j 2 .t 1.0 MeV) resulted in an irradiated average upper shelf energy decrease of 14 ft-lb, resulting;
- {
t in an irradiated upper shelf energy of 76 ft-lb. } l us .o The average upper shelf energy of the weld metal decreased 39 ft-lb after irradiation to I 4.040 x 10" n/cm (E > 1.0 MeV). His results in an irradiated upper shelf energy of 110 '[ 2 ft-lb for the. weld metal specimens. i o The average upper shelf energy of the weld HAZ metal &ed 22 ft-lb after irradiation to 4.040 x'10" n/cm (E > 1.0 MeV). This results in an irradiated upper shelf energy of 2 133 ft-lb for the weld HAZ metal. o l The surveillance Capsule W test results indicate that all measured 30 ft-lb transition temperature shift values are within a two sigma band of the Regulatory Guide 1.99, 0 Revision 2 3 30 ft-Ib transition temperature shift predictions. o he surveillance Capsule W test results indicate that the measured upper shelf energy . decreases of all surveillance materials are less than the Regulatory Guide 1.99, Revision 2. - predictions (Table 5-10). Additionally, all beltline materials exhibit a more than adequate - upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life of the vessel (32 EFPY) as required by 10CFR50, Appendix Grzi, o The calculated end-of-life (32 EFPY) maxirnum neutron fluence (E > 1.0 MeV) for the. Farley Unit I reactor vessel is as follows: Vessel inner radius * - 3.550 x 10" n/cm2 Vessel 1/4 thickness - 2.020 x 10" n/cm2 Vessel 3/4 thickness - 4.615 x 10" n/cm2
- Clad / base metal interface 1-2
r v: y:9 -a ~ m '0 - The 48 EFPY maximum neutron fluence (E > 1.0 MeV) for the Farley Unit I reactor vessel. ~ i is as followst Vessel' inner radius * - 5.194 x 10" n/cm' 2 Vessel 1/4 thickness - 2.955 x 10 n/cm - g. Vessel 3/4 thickness - - 6.752 x 10'8 n/cm T 2
- Clad / base metal interface ~.
J U P l-3
q r f .J v l 1 SECTION
2.0 INTRODUCTION
This repon presents the results of the examination of Capsule W, the fourth capsule to be removed . from the reactor in the continuing surveillance program.which monitors the effects of neutron j isradiation on the Alabama Power Company Farley Unit I reactor pressure vessel materials under. actual operating conditions. The surveillance program for the Farley Unit I reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program ~ and the preirradiation mechanical properties of the reactor vessel materials is presented in ' WCAP-8810, entitled "Southem Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.1 - Reactor Vessel Radiation Surveillance Program" by J. A. Davidson, et al.m. The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based on ASTM - E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Energy Systems personnel were contracted to aid in the preparation of procedures for removing Capsule W from the reactor and its shipment to the Westinghouse Science and Technology. Center Hot Cell Facility, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed. P This repoit summarizes the testing of and the postirradiation data obtained from surveillance Capsule i W removed from the Alabama Power Company Farley Unit I reactor vessel and discusses the analysis of the data. i ~ .ln o 2-1 j i
mm SECTION
3.0 BACKGROUND
ctor core and its primary coolant to h The ability of the large steel pressure vessel containing t e reauring safety in i bj cted to resist fracture constitutes an important factor n enst ritical region of the vessel because region of the reactor pressure vessel is the mos cll ffects of fast neutron irradiat significant fast neutron bombardment. The overa e h as SA 533 Grade B Class 1 (base mater l properties of low alloy, ferritic pressure vessel stee s suchell plate B6919-1) a of the Farley Unit I reactor pressure vessellower s i crease in hardness and tensile properties literature. Generally, low alloy ferritic materiais show an nt in conditions of irrad and a decrease in ductility and toughness under cer a in reactor pressure vessels has been f A method for performing analyses to guard against fast racture" Appendix presented in Protection Against Nonductile Failure,f cture mechanics conce and Pressure Vessel CodeM The method uses ra reference nil-ductility temperature (RTum). ility transition temperature (NDTT per d is defined as the greater of either the drop weight nil-uct50 ft-lb (and 3 RTum ASTM E-208m) or the temperature 60 F less than theimens with the long of a given materialis temperature as determined from Charpy spec of the plate. The RTym i (transverse orientation) to the major working direct onsity factor curve (K g curve) which appears in i i d static used to index that material to a reference stress ntencurve is a lower bound Appendix G to the ASME Code. The K a ure vessel steel. When a given material i f fracture toughness results obtained from several heats o press b i ed for this material as a curve, allowable stress intensity factors can be o ta nb determi f is indexed to the Km h function of temperature. Allowable operating limits can t en e stress intensity factors. lants can be adjusted to account for the and, in turn, the operating limits of nuclear power pi l roperties. The RTsm effects of radiation on the reactor vessel mater a p l steel can be monitored by a reactor mechanical properties of a given reactor pressure vesseReactor Vessel Rad surveillance program, such as the Farley Unit 1di ll removed from the h which a surveillance capsule is perio ca y h average Charpy V-notch 30 ft-lb temperature encapsulated specimens tested. The increase in t e 3-1 ~h~%
(ARTum) due to irradiation is added to the origin l RT embrittlement. This adjusted reference temperature (ART)um to adjus a index the material to the K curve and in tum to (ART = Initial RT um + ARTum) is used to ia which take into account the effects of irradiation on thset operating limits for t e reactor vessel materials. tF 3-2
l SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its primary coolant 6 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 irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 533 Grade B Class 1 (base material of the Farley Unit I reactor pressure vessel lower shell plate B6919-1) 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 under certain conditions of irradiation. A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code"1 The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTum). RT or is defined as the greater of either the drop weight nil-ductility transition temperature (NDIT per N ASTM E-208"l) or the temperature 60 F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens with the longitudinal axis oriented normal I (transverse orientation) to the major working direction of the plate. The RTuor of a given material is used to index that material to a reference stress intensity factor curve (K a curve) which appears in i Appendix G to the ASME Code. The K ai curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K,a 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. RTum 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 Farley Unit 1 Reactor Vessel Radiation Surveillance Program"l, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature 3-1
w h . (ARTym) due to irradiation is added to th original RTxur to adjust the RTxur for radiation s K ' embritdement.- This_ adjusted reference tespigure (ART) (ART = Initial RTuur + ART,m) is used to. index the material to the Km curve and, in turn, to set operating limits for the nuclear power plant. which take into account the effects of irradiation on the reactor vessel materials. z. l sn T% F i' .. t l i t i -h ) l r i s 3-2 1 e i [
SECTION 4.0 ' DESCRIFTION OF PROGRAM - Six surveillance capsules for monitoring the effects of neutron exposure on the Farley Unit I reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up. De six capsules were positioned in the reactor vessel between the neutron shielding puis and the - vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. Capsule W was removed after 12.43 Effective Full Power Years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact tension (CT) specimens (Figure 4-2) from a submerged are weldment fabricated with type B4 weld wire (heat number 33A277) and Linde 0091 flux (lot number 3922), and is representative of the intermediate shell longitudinal weld seams, and Charpy V-notch, tensile, CT, and bend bar specimens (Figure 4-2) from the lower shell plate B6919-1. Capsule W also contamed Charpy V-notch specimens from weld Heat-Affected-Zone (HAZ) material. All heat-affected-zone specimens were obtamed from within the HAZ of plate B6919-1. Test material obtained from the lower shell plate B6919-1 (after thermal heat treatment and formmg of - Ce plate) was taken from at least one plate thickness from the quenched ends of the plate. All test specimens were machined from the 1/4 thickness location of the plate after performing a simulated post-weld stress-relieving treatment on the test material. Test specimens were also removed fmm weld and heat-affected-zone metal of a stress-relieved weldment joining the lower shell plate B6919-1. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of the lower shell plate B6919-1. Charpy V-notch specimens from lower shell plate B6919-1 were machined in both the longitudinal orientation (longitudinal axis of the specimen parallel to the major working direction) and transverse orientation (longitudinal axis of the specimen perpendicular to the major working direction). Le Charpy V-notch specimens from the weld metal were machined such that the long dimension of the . specimen was normal to the weld direction. The notch was machined such that the direction of crack propagation is in the weld direction. 4-1
Tensile specimens from lower shell plate B6919-1 were machined with the longitudinal axis of the - specimens both normal and perpendicular to the major working direction of the plate. Weld specimens were oriented normal to the weld direction. Compact tension test specimens from lower shell plate B6919-1 were machined in both longitudinal and transverse orientations. Compact tension test specimens from the weld metal were machined normal to the weld direction with the notch oriented in the direction of the weld. All specimens were fatigue pre-cracked according to ASTM E399. Bend bar specimens were machined from plate B6919-1 with the longitudinal axis of the specimen oriented normal to the rolling direction of the plate such that the simulated crack would propagate in the rolling direction of the plate. All bend bar specimens were fatigue pre-cracked according to ASTM E399. The heat treatment and chemical composition of the surveillance materials is presented in Tables 4-1 and 4-2, respectively. Table 4-3 provides the Cu and Ni weight percent of the Farley Unit I reactor vessel beltline materials as reponed in Reference 6. Capsule W contained dosimeter wires of pure iron, copper, nickel, and aluminum 0.15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Neptunium (Np ") and Uranium (U ") used to measure the integrated flux at specific neutron energy 2 2 levels were included in the capsule. Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. These thermal monitors are used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two alloys and their melting points are: 2.5% Ag,97.5% Pb Melting Point 579 F (304 C) 1.75% Ag,0.75% Sn,97.5% Pb Melting Point 590 F (310* C) The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule W is shown in Figure 4-2. t 4-2 1
d. (i.-
- a.,,
1-J P J i . TABLE 4 ; m. -- Heat Treatment of the Parley Unit 1 Reetar Vessel Surveillance Mataials Material Tw.uua ('F) -Time (hours). Coolant Lower Shell-Austenitized 4
- Water quenched Plate B6919-1 1550 - 1650 ~
. Tempered 4 Air cooled: -1200 - 1250 Stress Relieved 40 Furnace cooled. 1125 - 1175 to 600*F Weld Metal - Stress Relieved 16 Furnace cooled. l 1125 - 1175 i? r i h f i e N i 4-3 J
TABLE '4-2 ' D3- ' Onirradiated Chemical Composition of the Parley Unit 1 Reactor Vessel Surveillance Materials - Element
- Lower Shell Plate B6919-1 W eld M etal Combustion Westinghouse Westinghouse' Analysis Engineering Analysis Analysis 0.13 C
0.20 S 0.015 0.013 0.009-N 0.003 0.005' 2 Co 0.008 0.016 0.018 Cu 0.14 0.10 0.14 Si 0.18 0.28 0.27 Mo 0.56 0.51 0.50 Ni 0.55 0.56 0.19 Mn 1.39 1.40-1.06 Cr 0.13 0.063 V <0.001 0.003 P 0.015 0.015 0.016 Sn 0.008 0.005 0.009 At 0.025 Nuits: All elements not listed are less than 0.010 weight percent. This unirradiated data was taken from the Farley Unit I surveillance capsule program. WCAP-8810. Reference 3. s f ~ T E 1 4-4 r
TABLE 4-3 W Cu and Ni Weight Percent of the Farley Unit 1 Reactor Vessel Beltline Materials Beltline Material Cu weight % Ni weight % Intermediate Shell Plate B6903-2 0.13 0.60 Intennediate Shell Plate B6903-3 0.12 0.56 Lower Shell Plate B6919 l-0.14 0.55 Lower Shell Plate B6919-2 0.14 0.56 Intermediate Shell Longitudinal Weld 0.25 0.21 Circumferential Weld 0.225 0.20 Lower Shell Longitudinal Weld 0.17 0.20 NUlt: This licensing basis data was taken from the Farley Response to GL 92-01. Reference 6. B 4 4-5
t 0 0 REACTOR VESSEL y Z CORE BARREL 1-NEUTRON PAD L___ V o I 270* goo U W i f I I IWAcTOR VESSEL NOTE: i CAPSULE IDENTIFICATIONS I HAVE CHANGED FROM THOSE IDENTIFIED IN WCAP-8810 l I i @f WALL VESSEL \\ s ) s' cAe8uus CORE ""'"" Y 'l111111llll I j's)' CORE f. l
- Q MIDPLANE i
l l s l 4 l s NElfmON PAD 2 l
- s f
" CORE BARREL f 1 ELEVATION VIEW Figure 4-1 Arrangement of Surveillance Capsules in the Farley Unit 1 Reactor Vessel 1 4-6 1 i
N l l SPECIMEN NUMBERING CODE: AT - PLATE B 6919-1 (TRANSVERSE ORIENTATION) AL - PLATE B 6919-1 (LONGITUDINAL ORIENTATION) i AW - CORE REGION WELD METAL AH - HEAT-AFFEpTED-ZONE METAL BEND BAR TENSILES COMPACTS COMPACTS CHARPYS CHARPYS CHARPYS COMPACTS COMPACTS CHARPYS CF AW9 AW45 AH45 AW42 AH42 AW3ll AH39 AW3G AH30 Ah W AT3 AV/8 AW12 AW11 AW10 AW9 AW44 AH44 AW41 AH41 AW3B AH38 AL12 AL11 AL10 ALS AW35 AH35 5 AW7 AW43 AH43 AW40 AH40 AW37 AFU7 AW34 AH34 A A vr Cu 'I ii Al.15%Co Cu
- 1 il 13 8'
II Il II Fe
- ;, gg g Fe wuw 6.-8 57.F
- -'s r 1 590*F r"1 G IT R l ll :
A1 15%Co (Cd) MONITOR si i l l! I I y i '.! !l 8 d 4 CEN TO TOP OF VESSEL +e., C~
T* P y NEUTRON SHIELD PAD / '# 5 l $ q *%' 1 ANSTEC y' =l APERTURE CAPSULE W l CGRE CARD ~ i Abo Ava!!able ors CORE BARREL Aperture Card / VESSEL. WALL CAPstiLE W Np'8' U"* JtPYS DOSIMETER TENSR.ES CHARPYS CHARPYS CHARPYS CHARPYS CHARPYS COMPACTS COMPACTS TENSILES 1 AH33
- ALS AT45 AL45 AT42 AL42 AT39 AL30 AT36 AL3O AT33 AL33 AT9 l AH32 194 AL3 AT44 AL44 AT41 AL41 AT38 AL30 AT35 AL35 AT32 AL32 AT12 AT11 AT10 AT9 ATS i AH31 ALy AT43 AL43 AT40 AL40 AT37 AL37 f.T34 AtJ4 AT31 AL31 A17 0
A k lr l Al.15%Co Ce l3 ! Al.15%Co k Is l le s j',iJ buU 9" Al.15%Co (Cd) 579 " f7 F"1 r"1 Al.15%Co (Cd) ll ' MONIT01 s Il I s 88 I I ll I i y8 N1 IIgI_l Ni .i m 8 iI i i ER REGION OF VESSEL TO BOTTOM OF VESSEL 8 Figure 4 2 Capsule W Diagram Showing Location of I Specimens, Thermal Monitors, and Dosimeters 9qo3/00 V27-0 J
w m y. s. SEC110N 5.0
- TESTING OF SPECIMENS FROM CAPSULE W L 5.1. Overview -
De post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at. the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Energy A and Hm,. Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G 5 ASTM Specification E185-82, and Westinghouse Remote Metallographic Fr.cility (RMF) Procedure 8402, Revision 2, as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1. 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-8810m,. 1 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). M and RMF Pmeedure - The Charpy impact tests were performed per ASTM Specification E23-93a 8103, Revision 1, on a Tinius-Olsen Model 74, 358J machine. De tup (stnker) of the Charpy. machine is instrumented with a GRC 830-I instrumentation system, feeding information,into an IBM compatible 486 computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurernent of Charpy energy (Eo). From the load-time curve (Appendix A),. the load of general yielding (Pay), the tirne to general yielding (tm), the maximum load (Pu), and the time to maximum load (tu) can be deternuned (see the Idealized load-time record, Figure A-1 in 2 Appendix A). Under some test conditions, a sharp drop in load indicative of fast fracture was 4-observed. The load at which fast fracture was initiated is identified as the fast fracture load (P,), and J-the load at which fast fracture terminated is identified as the arrest load (PJ. The energy at maximum load (Eu) was determmed by comparing the energy-time record and the load-time record. De energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Derefore, the propagation energy for the crack (E,,) is the difference between the total energy to fracture (En) and the energy at maximum load, (Eu). 5-1
The yield stress (cy) was calculated from the three-point bend formula having the following expression: L "'B(W-a)'C (1) where L = distance between the specimen supports in the impact testing 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 ($), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending). In three-point bending a Charpy specimen in which $ = 45* and p = 0.010", Equation I is valid with C = 1.21. Therefore (for L = 4W), L 3.33ParW oy-Pay B(W-a)2.21 B(W-a)2 1 (2) For the Charpy specimens, B = 0.394 in., W = 0.394 in., and a = 0.079 in. Equation 2 then reduces to: ay=33.3 Pay (3) where cy is in units of psi and Pay is in units of pounds. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula. Symbol A in columns 4,5, and 6 of Tables 5-5,5-6,5-7, and 5-8 is the cross-section area under the notch of the Charpy specimens: A-B(W-a)-0.1241 sq.in. (4) Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-92" i The lateral expansion was measured using a dial gage rig similar to that shown in the same specification. Tensile tests were performed on a 20,000-pound Instron Model 1115, split-console test machine, per ASTM Specification ES-93 "3 and E21-92n23, and RMF Procedure 8102, Revision 1. All pull rods, 1 grips, and pins were made ofInconel 718 hardened to 1IRC45. The upper pull rod was connected l 5-2
j g4 w V l thmugh a universaljoint toJmprove anality ofloadmg. De tests were cWM at a constant I crosshead speed of 0.05 inches per minute throughout the test. 4 a l m . Extension measurements were made with a linear variable displacement transducer extensometer. d 1 He extensometer knife edges were spring-loaded to the specimen and operated through specimen ' j failure. The extensometer gage length is 1.00 inch. De extensometer is rated as Class B-2 per l - ASTM E83-93 3"1, l 'I s Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with : ~ I - a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a j r thermocouple directly to the specimen, the following procedure was used to monitor ' specimen j temperature. Chromel-alumel thermocouples were inserted in shallow holes in the center and each ] - end of the gage section of a dummy specimen and in each grip. In the test configuration, with a. slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range of reorn temperature to 550*F (288'C). The - upper grip was used to control the furnace temperature. During the actual testmg the grip - temperatures were used to obtain desired specimen temperatures. Experiments indicated that this, r method is accurate to t 2*F. I ne yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined i directly from the load-extension curve. De yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. De final diameter and fmal gage length a I were deterndned from post-fracture photographs. De fracture area used to calculate the fracture i stress (true stress at fracture) and percent reduction in area was computed using the fmal diameter measurement. 1 5.2 Charny V-Notch Impact Test Results i i i 1 The results of the Charpy V-notch impact tests performed on the various materials contained in .l Capsule W. _which was irradiated to 4.040 x 10" n/cm (E > 1.0 MeV), are presented in Tables 5-1 2 i through 5-8 and are compared with unirradiated resultsm as shown in Figures 5-1 through 5-4.' The } - transition temperature increases and upper shelf energy decreases for the Capsule W materials are -{ summarized in Table 5-9. l L i 5-3 i i
h Irradiation of the reactor vessel lower shell plate B6919-1 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 4.040 x 10 n/cm (E > 1.0 MeV) (Figure 5-1) resulted in a 30 ft-lb transition 2 l temperature increase of 155*F and in a 50 ft-lb transition temperature increase of 160'F. This results in an irradiated 30 ft-lb transition temperature of 125 F and an irradiated 50 ft-lb transition temperature of 160'F. The average Upper Shelf Energy (USE) of the lower shell plate B6919-1 Charpy specimens (longitudinal orie itation) resulted in an energy decrease of 31 ft-lb after irradiation to 4.040 x 10 n/cm' (E > 1.0 MeV), This results in an irradiated average USE of 109 ft-Ib (Figure 5-1). Irradiation of the reactor vessel lower shell plate B6919-1 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse 2 orientation) to 4.040 x 10 n/cm (E > 1.0 MeV) (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 145'F and in a 50 ft-lb transition temperature increase of 135'F. This results in an irradiated 30 ft-lb transition temperature of 160 F and an irradiated 50 ft-lb transition temperature of 205 F. The average USE of the lower shell plate B6919-1 Charpy specimens (transverse orientation) 2 resulted in an energy decrease of 14 ft-lb after irradiation to 4.040 x 10 n/cm (E > 1.0 MeV). This resulted in an irradiated average USE of 76 ft-lb (Figure 5-2). Irradiation of the surveillance weld metal Charpy specimens to 4.040 x 10 n/cm (E > 1.0 MeV) 2 (Figure 5-3) resulted in a 95 F increase in 30 ft-Ib transition temperature and a 50 ft-lb transition temperature increase of 90*F. This resulted in an irradiated 30 ft lb transition temperature of 15 F and an irradiated 50 ft-lb transition temperature of 35*F. The average USE of the reactor vessel core region weld metal resulted in an energy decrease of 39 ft-lb after irradiation to 4.040 x 10 n/cm (E > 1.0 MeV). This results in an irradiated average USE 2 of 110 ft-lb (Figure 5-3). Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 4.040 x 10 2 n/cm (E > 1.0 MeV) (Figure 5-4) resulted in a 30 ft-lb transition temperature increase of 110 F and 5-4
) i i ' a 50 ft-Ib transition temperature increase of 120*F. This results in an irradiated 30 ft-lb transition , t.w ure of -40*F and an irradiated 50 ft-lb transition temperature of -5'F. i De average USE of the reactor vessel weld HAZ metal experienced an energy decrease of 22 ft-lb.' 2
- after irradiation to 4.040 x 10" n/cm (E > 1.0 MeV). His resulted in an irradiated average USE of 133 ft-lb (Figure 5-4).'
He fracture appearance of each irradiated Charpy specimen from the various materials is shown in o ' Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test l temperature. j l, A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the various Farley Unit I surveillance materials with predicted values using the methods of NRC i Regulatory Guide 1.99, Revision 2, is presented in Table 5-10. This comparison indicates that for ' the Capsule W surveillance materials: I ne measured 30 ft-lb transition temperature increase for the lower shell plate B6919-1 j longitudinally-oriented specimens is' 22'F higher than the Regulatory Guide 1.99, Revision 2, - 1 prediction. And, the measured 30 ft-lb transition temperature increase for the lower shell plate - B6919-1 transversely-oriented specimens is 12'F higher than the Regulatory Guide 1.99, - [ Revision 2, predictions. However, Regulatory Guide 1.99, Revision 2, requires that a two l sigma allowance of 34"F for base metal be added to the predicted reference transition j temperature to obtain a conservative upper bound value. Dus, the reference transition l temperature increases for lower shell plate B6919-1 are bounded by the two sigma allowance ) -i for shift prediction. De measured upper shelf energy decrease of lower shell plate B6919-1 is less than the Regulatory Guide 1.99, Revision 2, prediction. 1 The Capsule W surveillance weld metal measured 30 ft-lb transition temperature increase and j upper shelf energy decrease are less than the Regulatory Guide 1.99, Revision 2, predictions. j Re load-time records for individual instrumented Charpy specimen tests are shown in Appendix A. 5-5
5.3 Tension Test Results The results of the tension tests performed on the various materials contained in Capsule W irradiated to 4.040 x 10" n/cm (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated 2 resultsW as shown in Figures 5-9 through 5-11. The results of the tension tests performed on the lower shell plate B6919-1 (longitudinal orientation) 2 indicated that irradiation to 4.040 x 10" n/cm (E > 1.0 MeV) caused less than a 19 ksi increase in the 0.2 percent offset yield strength and less than a 18 ksi increase in the ultimate tensile strength when compared to unirradiated datam (Figure 5-9). The results of the tension tests performed on the lower shell plate B6919-1 (transverse orientation) 2 indicated that irradiation to 4.040 x 10" n/cm (E > 1.0 MeV) caused less than a 18 ksi increase in the 0.2 percent offset yield strength and less than a 15 ksi increase in the ultimate tensile strength when compared to unirradiated dataW (Figure 5-10). The results of the tension tests performed on the surveillance weld metal indicated that irradiation to 4.040 x 10" n/cm (E > 1.0 MeV) caused less than a 12 ksi increase in the 0.2 percent offset yield 2 strength and less than a 9 ksi increase in the ultimate tensile strength when conipared to unirradiated datam (Figure 5-11). l The fractured tension specimens for the lower shell plate B6919-1 material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14. I The engineering stress-strain curves for the tensile tests are shown in Figures 5-15 through 5-20. 5.4 Compact Tension Tests J Per the surveillance capsule testing contract with the Southern Nuclear Operating Company, the bend bar and 1/2-T compact tension fracture mer.hanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center Hot Cell. 5-6 1
t . TABLE 5-1 -- ' Charpy V-notch Impact Data for the Farley Unit 1 Iower Shell Plate B6919-l= (imgitudinal Onentation) hradiated to a Fluence of 4.040 x 10 n/cm (E > 1.0 MeV) 8 2 a i Sameple Temperatum Impact Energy Imral Expansion Shear j Number T 'C ft-ibs Joules mits nun AIAI 25 -4 4 5 2 0.05 i AL40 75 24 24 33 20 0.51 5 AIA2 100 38 27 37 23 0.58 8 AIA3 110 43 14 19 11 0.28 15 AL32 125 52 31 42 23 0.58 15 AL34 140 60 20 27 22-0.56 - 20 AL37 150 66 77 104 58 1.47 45 . AL31 175 79 48 65 38 0.97 25 l AL33 200 93 53 72 41 1.04 60 AL44 207 97 98 133 72 1.83 75 AL39 225 107 95 129 72 1.83 95 AL38 250 121 99 134 73 1.85 90 AIAS 275 135 106 144 77 1.% - 1(X) AL36 300 149 103 140 78 1.98 1(X) AL35 350 177 117 159 84 2.13 l(X) ? r 9 f 5-7 4
~ ,a.. i TABLE 5 Charpy V-notch Impact Data for the Farley Unit 1 Iower Shell Plate B6919-1 (Transverse Orientation) Irradiated to a Fluence of 4.040 x 10 n/cm (E > 1.0 MeV) 2 I b Sample Temperature - Impact Energy Lateral Fwpandam Sheer Number T
- C ft-Ibs Joules mBs num AT44 75 24 14 19 11 0.28 5
AT45 100 38 17 23 14 - 0.36 i 10 AT37 125 52 29 39 26 0.66 15 AT36 150 66 28 38 22 0.56 15 AT32 160 71 34 46 26 0.66 25 AT43 175 79 36 49 29 0.74 25 AT39 175 79 24 33 24 0.61-40 AT33 185 85 45 61 29 0.74 35 AT41 200 93 40 54 35 0.89 45 AT31 225 107 50 68 44 1.12 90 AT34 250 121 68 92 57 1.45 100 AT40 275 135 78 106 65 1.65 100 AT38 300 149 75 102 60 1.52 100 AT42 325 163 78 106 56 1.42 100 AT35 350 177 79 107 68 1.73 100 l p i 5-8
m ' TABLE 5-3 Charpy V-notch Impact Data for the Parley Unit 1 Surveillance Weld Metal 2 Irradiated to a Fluence of 4.040 x 10 n/cm (E > 1.0 MeV) Sample Temperature Impact Energy Lateral Expansion Shear Number 'F l
- C ft-Ibs Joules mits l
nun AW45 -25 32 3 4 I 0.03 10 AW42 0 -18 12 16 9 0.23 15 AW34 0 18 8 10 0.25 15 AW36 10 -12 26 35 20 0.51 20 AW41 25 -4 42 57 30 0.76 25 AW33 25 -4 17 23 14 0.36 25 AW35 35 2 48 65 37 0.94 30 AW37 50 10 64 87 45 1.14 65 AW31 100 38 83 113 57 1.45 80 AW32 150 66 97 132 71 1.80 90 AW43 200 93 101 137 77 1.96 100 AW40 225 107 115 156 76 1.93 100 AW39 250 121 118 160 82 2.08 100 AW38 300 149 113 153 83 2.11 100 AW44 350 177 105 142 78 1.98 100 I \\ 5-9
TABLE 5-4 Charpy V-notch Impact Data for the Parley Unit 1 Heat-Affected-Zone (HAZ) Metal 2 hradiated to a Fluence of 4.040 x 10 n/cm (E > 1.0 MeV) s i 1 Sample Temperature Impact Energy Lateral Expanske Shear Number T
- C ft lbs l Joules mils mm AH38
-75 59 21 28 13 0.33 5 L AH40 -50 -46 25 34 16 0.41 10 l AH37 -40 -40 23 31 18 0.46 15 AH36 25 -32 50 68 27 - 0.69 25 AH32 -25 -32 32 43 17 0.43 15 AH44 0 -18 55 75 25 0.64 25 AH33 25 -4 65 88 36 0.91-40 AH43 50 10 82 111 53 1.35 35 AH42 100 38 101 137 65 1.65 80 AH45 125 52 126 171 75 1.91 90 AH31 150 66 143 194 74 1.88 K)0 AH35 175 79 130 176 77 1.% 100 AH34 200 93 130 176 76 1.93 100 AH39 250 121 135 183 68 1.73 IrX) AH41 300 149 128 174 90 2.29 100 5 1 t T 5-10
TABLE 5-5 Instrumented Charpy Impact Test Results for the Farley Unit 1 IAwer Shell Plate B6919-1 (Iengitudinal Orientation) Irradiated to a Fluence of 4.040 x 10" n/cm (E > 1.0 MeV) 2 Normaltrod Energies Sompie Test Charpy ft-Ibs/In Yleed Dme to Max. Time to Fracture Anoet Yletd. Flow Number Term Energy Charpy Max. Prop. Lood YleId Load Max. Load lead Strees Strees (*F) (ft the) Ed/A Em/A Ep/A Ubs) (mecc) Obe) (mese) Obs) Obs) (tal) (ksi) AIAI 25 4 32 16 16 2149 0.11 2149 0.11 2149 51 71 71 AIA0 75 24 193 137 56 3371 0.15 4337 0.35 4253 281 112 128 AIA2 100 27 217 169 48 3469 0.15 4503 0.41 4503 213 115 132 A1A3 110 14 113 40 72 3421 0.19 3421 0.19 3421 1007 114 114 AL32 125 31 250 185 64 3364 0.14 4508 0.44 4508 405 112 131 AL34 140 20 161 72 89 3291 0.2 3696 0.28 3696 1149 109 116 AL37 150 77 620 266 354 3387 0.14 4703 0.57 4097 1387 113 134 y AL31 175 48 387 282 105 3331 0.14 4595 0.6 4595 1667 Ill 132 O AL33 200 53 427 258 169 3393 0.17 4377 0.59 4377 2128 113 129 AL44 207 98 789 322 467 3271 0.14 4552 0.69 297I 1530 109 130 AL39 225 95 765 306 459 3089 0.14 4386 0 67 2092 1629 103 124 AL38 250 99 797 258 540 3069 0.21 4415 0.64 1713 951 102 124 AIA5 275 106 854 314 540 2949 0.14 43 % 0.69 N/A N/A 98 122 AL36 300 103 829 306 523 3012 0.14 4365 0 68 N/A N/A 100 123 AL35 350 117 942 314 628 3028 0.14 4366 0.69 N/A N/A 101 123 N/A - Fully ductile fracture. No arrest load.
TABLE 5-6 Instrumented Charpy Impact Test Results for the Farley Unit 1 I2)wer Shell Plate B6919-1 (Transverse Orientation) 2 Irradiated to a Fluence of 4.040 x 10 n/cm (E > 1.0 MeV) s6ormonzed Energies Sample Test Chorpy ft-Ibs/in YleM Time to Max. Time to Fracture Arvost YleM Flow Number Temp Energy Chorpy Max. Psop. Load Yleki Load Max. Lood Load Strees Strees (*F) (ft-R>s) Ee/A Em/A Ep/A (Ras) (meoc) (1bs) (meec) (Itas) (Ibs) (ksi) (kel) AT44 75 14 113 64 - 48 3614 0.15 4014 0.21 4014 280 120 127 AT45 100 17 137 81 56 3504 0.14 3990 0.25 3990 703 116 124 AT37 125 29 234 137 97 3475 0.16 4261 0.36 4261 1055 115 128 AT36 150 28 225 153 72 3381 0.14 4312 0.38 4311 1063 112 128 AT32 160 34 274 177 97 3361 0.14 4406 0.42 4406 1784 112 129 AT43 175 36 290 169 121 3238 0.14 4274 0.42 4274 1%3 108 125 AT39 175 24 193 32 161 3052 0.14 3052 0.14 3052 2083 101 101 AT33 185 45 362 234 129 3400 0.2 4537 0.56_ 4520 1972 113 132 AT41 200 40 322 177 145 3140 0.14 4187 0.44 4187 2226 104 122 AT31 225 50 403 177 225 3154 0.14 4199 0.44 4148 1117 105 122 AT34 250 68 548 209 338 3124 0.14 4202 0.51 N/A N/A 104 122 AT40 275 78 628 217 411 3076 0.16 4197 0.53 N/A N/A 102 121 AT38 300 75 604 209 395 3040 0.16 4124 0.51 N/A N/A 101 119 AT42 325 78 628 209 419 2801 0.14 4059 0.57 N/A N/A 93 114 AT35 350 79 636 225 411 3127 0.14 4331 0.52 N/A N/A 104 124 N/A - Fully ductile fracture. No arrest load. e ~. -, .v.v- -.r.- ,, - - -, - -. _ +, -....,.,, _,
1 l l TABLE 5-7 Instmmented Charpy Impact Test Results for the Farley Unit 1 Surveillance Weld Metal 2 l Irradiated to a Fluence of 4.040 x 10" n/cm (E > 1.0 MeV) l l Normailms Enogies l Somplo Test Chowpy ft-ts/In Yleid Time to Mox. Time to Feocture Anest Yloki Flow Number Temp Energy Charpy Max. Prop. Lood Ylokf Load Max. Lood Lood Strees Strees (*F) Ut-ibs) Ee/A Em/A E,/A Obs) (mooc) Obs) (meec) Obe) Obs) (ksi) (tal) AW45 -25 3 24 12 12 1940 0.11 1940 0.11 1930 0 64 64 AW42 0 12 97 48 48 4082 0.19 4082 0.19 4082 382 136 136 AW34 0 8 64 32 32 3378 0.14 3378 0.14 3378 315 112 112 AW36 10 26 209 137 72 3852 0.15 4604 0.33 4604 817 128 140 AW41 25 42 338 250 89 3739 0.15 4728 0.52 4728 1483 124 141 AW33 25 17 137 48 89 3775 0.15 3939 0.17 3939 1278 125 128 AW35 35 48 387 266 121 3692 0.14 4743 0.54 4729 1723 123 140 AW37 50 64 515 250 266 3877 0.16 4757 0.52 4695 2759 129 143 AW31 100 83 668 242 427 3622 0.15 4675 0.52 4261 3187 120 138 AW32 150 97 781 242 540 3409 0.15 4492 0.54 3058 2156 113 131 AW43 200 101 813 306 507 3178 0.14 4228 0.68 N/A N/A 106 123 AW40 225 115 926 298 628 3115 0.14 4270 0.68 N/A N/A 103 123 AW39 25) I18 950 298 652 3046 0.14 4170 0.68 N/A N/A 101 120 AW38 30) 113 910 290 620 2910 0.14 4073 0.69 N/A N/A 97 116 AW44 350 105 845 298 548 3080 0.14 4204 0.68 N/A N/A 102 121 N/A - Fully ductile fracture. No arrest load.
TABLE 5-8 I Instrumented Charpy Impact Test Results for the Farley Unit i Heat-Affected-Zone (HAZ) Metal 2 Irradiated to a Fluence of 4.040 x 10" n/cm (E > 1.0 MeV) l Normanned EnerWes I sormie Test chorpy ft.Ibs/in' Yleed 2no to Mon. Timele Feoctwo Anest Yleed Flow Munter form Eneggy Chorpy Mou. Prop. Lood Yleed Lood Mon. Lood Lood SIrees 38nses (*F) Ut-Ibs) Ee/A Em/A Ep/A Ubs) (mese) Obs) (mecc) Obs) Obs) (ks0 (ks0 AH38 -75 21 169 129 40 4843 0.19 5258 0.3 5258 70 161' 168 AH40 -50 25 201 161 40 4597 0.17 5118 0.34 5118 58 153 161 AH37 -40 23 185 137 48 4477 0.19 5046 0.32 5046 55 149 158 AH36 -25 50 403 290 113 4499 0.17 5324 0.54 5307 1452 149 163 AH32 -25 32 258 177 81 4583 0.18 5147 0.37 5147 683 152 162 AH44 0 55 443 266 177 4043 0.15 5134 0.52 4967 450 134-152 AH33 25 65 523 314 209 4339 0.2 5235 0.61 5167 1936 144 159 tp AH43 50 82 660 306 354 4186 0.16 5237 0.57 4760 2069 139 156 AH42 100 101 813 290 523 3988 0.17 4987 0.59 3472 2388 132 149 AH45 125 126 1015 354 660 3703 0.14 5047 0.68 2927 1947 123 145 All31 150 143 1151 346 805 3743 0.16 4914 0.69 N/A N/A 124 144 AH35 175 130 1047 370 676 3518 0.19 4849 0.75 N/A N/A 117 139 All34 200 130 1047 346 701 3536 0.14 4930 0.69 N/A N/A 117 141 AH39 250 135 1087 346 741 3431 0.16 4750 0.71 N/A N/A 114 136 AH41 300 128 1031 330 701 3529 0.16 4663 0.69 N/A N/A 117 136 N/A - Fully ductile fracture. No arrest load.
7 t : -- TABLE 5-9 Effect of Irradiation to 4.040 x 10" n/cm' (E > 1.0 MeV) on the Notch Toughness Properties of the Farley Unit i Reactor Vessel Surveillance Materials Average 30 ft-Ib (a) Average 35 udi(e) Average SS ft-h'* Average Upper Sint Energy Transitten Tesaperature (*F) Lateral Esponsion Temperatore (*7) Transition Tesaperature (*F) . tft-Ib). Meterial ~ Unirradisied Irradiated AT Unteradisied Irradiated AT Unirradiesed Irradiated AT Unirredlesed Irradiated AE Nte B6919-1 -30 125 155 -10 150. 160 0 160 160 140 109 - 31. (Imgitudinal) mte B6919-1 15 160 145 45 200 155 70 205 135 90 ~76 14 (Transverse) Weld Metal -80 15 95 -50 ' 30 80 -55 35-90 149 -. I10 - 39 lIAZ Metal -150 -40 110 -105 20 125 -125 -5 120 155 133 22 (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).
TABLE 5-10 Comparison of the Farley 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-Ib Transithm Upper SheK En-rgy Temperature Shift Decrease Huence Predicted "' Measured Predicted "' Measured (10" n/cm', E>1.0MeV) ('F) (*F) (%) (%) Plate B6919-1 Y 0.6014 84 85 20 9 (Longitudinal) U 1.817 114 105 26 21 X 3.062 127 135 30 19 W 4.040 133 155 32 22 Plate B69191 Y 0.6014 84 55 20 0 U l.817 114 90 26 9 X 3.062 127 105 30 11 y W 4.040 133 145 32 16 Weld Metal Y 0.6014 109 80 35 13 U l.817 148 80 45 28 X 3.062 165 100 50 23 W 4.040 173 95 53 26 HAZ Metal Y 0.6014 60 11 120 26 U l.817 125 19 X 3.062 W 4.040 110 14 (a) Based on Regulatory Guide 1.99, Revismn 2, methodology.
TABLE 5-11 Tensile Properties for the Parley Unit 1 Reactor Vessel Surveillance Materials Irradiated to 4.040 x 10 n/cm' (E > 1.0 MeV) Material Semple Test Temp 8.2% Ylehl Ultimate Fracture Fracture Fracture Uniform Total Reduction Number ('F) Strength Strength Lead Stress Strength E"_ _"; Elongation in Area ^ (ks0 (ks0 (kip) (ksi) (ke0 (%) (%) (%) Lower Shell AL7 150 81.5 98A 3.20 169.6 65.2 10.5 32.1 62 Plate B6919-1 AL8 225 78.9 95.7 3.25 183.9 66.2 9.8 22.1 64 Omnitudina0 AL9 550 73.3 93.1 3.50 158.8 71.3 9.5 18 5 55 Lower Shell AT7 200 78.9 95.1 3.55 139.5 723 9.8 18.6 48 Plate B6919-1 AT8 250 77.9 943 3.58 144.7 72.9 10.5 19.0 50 (Transverse) AT9 550 72.1 933 3.70 117.8 75A 9.0 15.6 36 AW7 75 89.6 96.8 3.10 187.7 63.2 9.8 23.7 66 Weld AW8 200 85.1 93.7 3.00 194.9 61.1 9.0 21.6 69 AW9 550 74A 91.7 3.05 172.6 62.1 9.8 22.5 64 YG
y-f I (*C) -150 -100 -50 0 50 100 150 200 250-c 1 I I I I I I I I 100 A 8 80 % 60 40 8 ts 't 20 I I I I I I 0 100 2.5 E e J n j 80 e 2.0 g 60 e 1.5 " 40 0 m e 1.0 I I I I I O O 160 140 ^ se 0 120 160 2 100 7 d 120
- 80 g
v b o @ 60 80 m m a ~ 12 40 l l l l l l 0 0 -200 -100 0 100 200 300 400 500 TEMPERATURE ('D tas O IBERAMATO e mRAMA10 cisin IUDE ua x lo"n/m a:) u ww t Figure 5-1 Charpy V-Notch Impact Propenies f'or Farley Unit 1 Reactor Vessel lower Shell Plate B6919-1 (longitudinal Orientation) 5-18
G: ('C) -150 -100 -50 0 50 100 150 200 250 I l l 1 l l l l l ^'3 100 O 8 80 60 0 6+ 40 as 20 I I I I I I 0 100 25 ^m3 80 en e v g 60 w 1.5 e " 40 m 1.0 $ 20 o 05 I I I I I I I 0 O 120 160 100 9 m 120 l 2 80 O 3 3 b D 60 2' 80 G m 40 ,e O e 40 0 O O 20 0 I I I I I I I 0 O -200 -100 0 100 200 300 400 500-TEMPERATURE (*F) ause o UglutARIATO U 2 e an@lATO W4 FUDCE 4HO x N n/m E ) U MeV) l Figure 5-2 Charpy V-Notch Impact Propenies for Farley Unit 1 Reactor Vessel Imver Shell Plate B6919-1 (Transverse Orientation) 5-19 i
g-p F ('C) -150 -100 -50 0 50 100 150 200 250 I I I I 2I 3 _ z2_ }_I I I s 1 100 V e f 8 80 60
- 40 o
20 I I I I I I I 0 100 2.5 a
- 80 6
F., " 2.0 e t5 g60 1.0 o e 3 40 3 20 es 0.5 I I I I I 0 0 200 180 240 160 o le ^ 200 140 o 2 ~ 120 3 o ,e 160 O D 100 e ~ 80 o 60 80 40 ~ 20 of ], I 6 .I I I I I o o -200 -100 0 100 200 300 400 500 TEMPERATURE ('D vtu KAH o ugRAMATD 8 e MANATO m(n rucct ua 2 a/mtE > u kW Figure 5-3 Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Surveillance Weld Metal 5-20
e (*O l -150 -100 -50 0 50 100 150 200 250 I I I l-1 I I I i ^ ^ ^^'3^ ^ 100 ^^ 8 80 'o o !E 60 o 40 20 7 I I I I I I I 0 100 2.5 3 80 o - ^5 24 6 e 60 o M 40 2 1.0 0 l2s 0 0 200 180 240 160 O g 140 200 g m 3120 160 6 b e 100 ~ 80 60 o 80 a 40 / o,[ 40 20 / eA 8 V I I I I I i g o -200 -100 0 100 200 300 400 500 TEMPERATUPI ('D w Nm o LN151AIRAD e siisAs cise*n FLton 4m x #va E > u mn t Figure 5-4 Charpy V-Notch Impact Propenics for Farley Unit 1 Reactor Vessel Weld Heat-Affected-Zone (HAZ) Metal i 5-21 I
l1 2,3 ~ ~ i
- p -
- 3. s,5;
.j M: ,f I ih 3.47, ,.4 c,(..' ,r . ma. .i g g.'j-i' [.' eQ q - " riMt 1. f. y + 4 e n,:g ?n t i
- L U
+ !'n t r ( ff i U f. .$s.,..,p; t. 9, p' A ~. ;, a q LW xm:o)h p _1: cr:39
- i. -
t w.- -,s AL41 AL40 AL42 AL43 ,AL32, s. ~, - ~, ,~y ~, [L [ l; 3 i f C; } 'f (. [f-a p b 4 i. 1 v-a s; tL s 1.' N i 1 ', i.. ..3 v;, ;.- g [W G \\_; L [ i.: 'ye -e{~- r:.s _-.:.-~. AL37 _,, AL31, _...., , AL33.,.y,_, _,,, AL44 AL34_. 7-~*.. y ~ ~ y, p.- g .t . 3 y 4 Ih ~ ,A r, c l g~- a 7 - 13 g h: j f,)<- ~ i h- .. /.] a ,i p y } 'i ll ~ a 1 L U i ~ - ^ ~~' AL38 AL45 AL36 AL35 ~ l 1 t Figure 5-5 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Reactor Vessel Lower l ShcIl Plate B6919-1 (longitudinal Orientation) i .l ) l 5-22 i
4
- l
- j
,..f 2- ,c,-. 3;.u:,s '~- ^ i, f-I- l .,m. r f p ' ' t}, (,' :.t we ..e ,j;, j, [f, j. 7-3 f Y
- 3 u
l i 4 j: j. >4
- 1..
i M [ h p b' 3 N' 'p - n + [f } YW5, y l ~; i g i .:.& w ,...... ; 2, u L _... .a,, L. ? AT43 AT39 AT33 AT41 AT31 n... -~ c.. m -.,.,, ,-w-n-
- 7., -
n-e, n,, en,,-- lC * , (;
- 4 i.
!f;
- fb V
i ) f..'. 4 , Jf ,,i- ~ 'l I. L ad e 1 i ~, i L. nd, + ?O .r w ~ i i Figure 5-6 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Reactor Vessel Lower i Shell Plate B6919-1 (Transverse Orientation) 5-23
- ~
7 ..n m ;m:w.w :r m m.- , m - - w -r mm ~-n. m 4 5< e ;. ~ 4 y. i;[ h 3 Jl l:'1 '9 i se l ,k a-t a. c i: ~- .f[ b'* 'i [ e. j f ,g e. e-. g a t ;c ,e
- e..
,-n x e mA --- : aJw-maa
- e. c a,- _
.s o_ m.. AW34 AW36 ~,2 AW4 m.m . AW42. n.
- AW45,
- g y +.e,,g a - c
- ner, u
n , ;,. c '.u. m.- -(.. k' f
- s. ]
i f: C t: C \\] p 4; g sc -- h q q ( I' ? l j, f.; g {' q ( g; p l] U i +- 4 p'l ng t 3 vj 1 4 .,t 4 ? { 1"., g;g 1-1 V' # 1 e< p tl Y.' \\, a l ".;. 4] y j l. lj 4 ^ j A .__.a i _ awn QU \\, n, a u .] .m r
- 7..m.m.-,,3_3,,,,.., AW35
,,A_W37y,,,,,m.,.7 AW31.,y,,y.y,A._W3_2 _,y, AW E ., {y ^ & vI.
- 9:p q.
p. [,'t y, y
- 4 -
b, p ^ f ;.:, t i b , g, I,:.< d c (
- g s
c
- 5
[-? f: 'l} 'Y;, ,r J y i; 3;. V E d;' .. + X. b l 'Y T. ;.., L...i - g.--~. J '. J,. i ;t .~ }" M i - %e ,. ~, I I i l I I Figure 5-7 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Reactor Vessel l Surveillance Weld Metal i l 5-24
g,m.,-,,.- ~ ' ". A ; }- gaypye; 3/WG f: M4~ . z 491. l- ) syg. ];; g V l j :. g , p a t Y li & { l. l) 1 AH38 AH40 A,H37 AH36 .AH32 t ~ ~ c-x f f 4 o j l [ l f i \\; l ). o cc 3;;$a ? _.. m ~_m,.. A. & am;!7,. $n { AH.4% _,, AH33.,,,,,,_AH4.3,., AH42.,.., , AH45 y., < a, -Q l3 l), [ l. l'
- )
. -) 'im i s o .1 ^ l
- t
'n e, i .;5.. ' l Li ^ s a t -l J ~ AH31 kH35 AH34 AH39 AH41 Figure 5-8 Charpy Impact Specimen Fractusc Surfaces for Farley Unit 1 Reactor Vessel Weld Heat-Affected-Zone (HAZ) Metal 5-25
-r ( l cc) 0' 50 100 150. 200 250 300 l '120 g j g g l l l-800 110 700 100-9. 90 ga m/. { 600 Na t% n .h i u 80 e s t e h! O- - 500 M 70 azzYnn st o Gm ' 400 60 - 8 s, I I I I I ~ 40 80 I l eDucfm mAKA C 70 g 60 8 50 U3 40 EAL DDGTm U30 gk ,,e ,f t 20 2 10 - e h-S 0 0 uersn assim r l I I I 0 0 100 200 300 400 500 600 TEMPERATURE (*F) tas A O leenRADIATO [ 0 A e MARIATO AT 550'F, FLDCE 4340 x 10 nhm2 E )10NrW l 1 l l Figure 5-9 Tensile Properties for Farley Unit 1 Reactor Vessel lower Shell Plate B6919-1 s (longitudinal Orientation) I s 5-26 i ) 2 s
c+- .(... Fl f I (*C) 0 50~ 100-150 200 250 300-120 l l-l l l l l-800 110 700-100 1 i i ?,i 90 A u.TNTE TDERI ITEDsm 600 6 A e 500
- 70
' 60 a o o v-400 50 i i l I I 300 4g f 80 70 l 60 8 50 C3 40 30 "" Y " t l A 20 ^ ^ I " ~ t o i 10 O E "ueran amEAta 1 I I I I 0 0 100 200 300 400 500 600 TEMPERATURE (*D vas i inus A O usRAIEAfD A 91RRAIRATO At 5507,FLBCE 040 x la8a/cn2 E ) to neV) I a Figure 5-10 Tensile Properties for Farley Unit 1 Reactor Vessel lower Shell Plate B6919-1 (Transverse Orientation) 5-27 6 m-. m
n 1 '('C) 0 50 100 150 200 250 300 1 800 i 120 110 L-700 100 lLTIMATE TDGlu: :TEDETH e 4 s i 9 90. Q ]- 600 g 6 e ,p M 80 k W e 500 G 70 n a vr a :Tu r ra 60 400 j 50 l 300 l l l l l 40 .t l. 80 notcfm IN MA 70 C ga ( 60 8 50 C 40 cd $ 30 TDTAL DDEATM ^,e o i 20 i 10 O E E' usumitarain l I I I O 0 100 200 300 400 500 600 ~, TEMPERATURE (*D ALA87 A O taERRAllATO f A G BREATO AT 527, FUDCE 4840 x IO a/cn2 g > 13 ney) { U t r r l Figure 5-11 Tensile Propenies for Farley Unit 1 Reactor Vessel Surveillance Weld Metal i '.l I 5-28 l 9
I 1 i i w w m,,m,.mm.,-mn+m;v... v.a,mwww. -
- t. ~.
a v. .. g ..,.. u. .p ; j..[l. <... ~ yg c:.p},p o '
- .p$ -
a 4., v.. 4 .. w., g n g4m, r.gij r / p / 3 a ;t. N > ~,i 6 s 4o .n - m 7 ; 4 b; $ ~ ' ' ' ' 1 'u .i' t. . - j, 4 31 P ,.f hi p A 4'.,,. f N p 5, '. ' 7 V.( .'; r'; n 't.N,' ) ;ij J ',939 e 4 h( d, f."r,
- ne %
""e ,., l: 1
- P s y_:q g_Q 1.
gj. h ', w...' r i d, . m "%c-~r ' 's.'t OrtaLk:mssaan;/y;hf. ' tF n >.w. -.~ cm Specimen AL7 Tested at 1500F I mwa'lm% %< pe '- - -a.At % r.e. 5 6.5,;..m; m ,.,vu.e *$~. m,,.- m-m n c s ._j . s ': n V 4 YE5 kf lhNfhhhj.h}l hf(( S a.a.. x v -. 4 I i I, i N.' h,i 4 m WQ J ;h + Q.., XC: -i!&t n. 1"g:'yy :s Lf -W y^. f i c p.-3 ;
- p/ s 4
u + w.; e,.ap'; \\q \\geh. m m ~ ;,_ {'>1f. J kbjl,b[. $' _ $' m W' q.;&' lf k,hK,':t ..n .W l ~ ,1,p : o y ",0 M g-e c. , _,Q' ju.Q,y,.Q,2 t_V
- 7'4.;;.
a ltq \\ l (m.4 : wd W.. m, W. r gw w c i y a n S M bM D a d E d i s E;sdd d s M a l b. '
- o s
Specimen AL8 Tested at 225'F i i p m e":= m a n.p m m g e y p m w a r y - %% ktQ2lll#FJ4fyfM f jb c ' y W G yNGlyQQ fR l 1,, c a I. 1 r + l L 2 t m... ~ ~ %4& ,..'e..*. 1 W;....Wl., d * [2 ;i;i P. g t s i , ; yi.. VL C, 4 _<., ,v. , ?!s i ..e b iw
- q.
n4_"' g a:' ,7l' ~v s f !.y;,:j ]> } __ _ q,_*. r
- y.
_;,._,, g'Q,:,\\ v \\ .; p N:h0.%'OAM$dah L;,,3$5&h.hilih0EN Ud k 15 1 Specimen AL9 Tes*ed at 550oF j ? l i i 4 h Figure 5-12 Fractured Tensile Specimens from Parley Unit 1 Reactor Vessel Lower Shell Plate t i i B6919-1 (longitudinal Orientation) l i t 6 i 5-29 i I I l l i ., _ _. _ _ _ _... _ _ _ _ _. f
F p.- ~.,_ .g 3.)y._,7q7, .q p. .g y n :;r;:w~ m. 3', , g ;w.. - uw - a y.m ~: 3 s q v. y
- _3,3;;S?i wgp 4lIlk hiM<z;.;,.
j' 'b ' s f h. [d( h . 3 4/ 7 ~ ) i 'O noJn
- f@f,t.p> A M,. +.c.Q
~, 1
- o.,
, ' ' l,' i l: ..,n, -< r s .:a. =+c,a .? ti (.y ) n + a,,. T, OA' l:;. ~:^.uawtum u.:,;.;A L %.&n kd% Wu.,a m is k; &n kunwMaAnuikkuu i Specimen AT7 Tested at 2000F. dA ?1 -g e. ~ n ih.' ;:;:g'? N "6 ' l:l[.wem' 1 .,99 p R{ L ^$W fYlll gQ g i g gwoy
- p 4;
.;. y:}, c p) ;;g.'s ' e ' Q i:+ ' & ' r hW
- [.7 [g ;ys Qgni a
',t f,', 5r o i@j ..,,e.~ :yr..g arsm e, ~,, . a. .4 r. m~yt gg j;.p&-m yliG;y%s 4 ',Q a: p' g <* Sr., ne ny. r,u. t. l;gj %ylty-y,.,g-rpjH,/ art.
- x tr
%t M mnpn ; 4BM c C C7 W QdQQ Specimen AT8 Tested at 2500F ..t l = [*MW37**fg*FM"writy c%yny'P,muyJg%%#tugimmy ,.;g 7 - ;c
- x., m.y
-,% s <g., o,s, -,.p:/c, - 9.#-...w +- e g s - -.-s g ,w I /4 . 5: 9 ;; x I 3$ 4, a s
- f. " ;r p +.v
%; ^ - _ ' -,y y_;
- ;,2.-
wij + ' ,o r r- : u
- < t ;2 rr >
5 + t E ) 4 d I I 7.,, - h 2 7,x t.- - : x: v.
- .ll t - Iy(,'a
- i.G;y,,ps.
- s.,.,
x, - e s e Jr. '8 Q v r .s 'g"j w .. g. j i m i,,, r? p r t mLem,a,s.az u ss %,sas. b s r a i p Specimen AT9 Tested at 550aF 1 1 \\ l l l Figure 5-13 Fractured Tensile Specimens from Farley Unit 1 Reactor Vessel Iower Shell Plate B6919-1 (Transverse Orientation) 4 l 4 l 5-30 1 1 l i l
J 4 l MIST 8QmTW"M@'"w YW7M. g'T. ~T?$?""'*E N'S&*'FN"M. a _h.' (w l k., .+g n f;.~ N c-Rl0 7 -Q ,QI _.,.,(, 3
- 4,,. qp y" 3-7
( y 4 ., n r s r 1 g y -4 i m %;p;hp.e f(t:[IID. iL- 't d;w ice dh
- s, u pymms u m gu.
a e my - y jyrgy+ y. y i;gggMqg, , m_
- y__,
~ m .~ o Specimen AW7 Tested at 75'F am-:nmetrxm-r-- glar myiigg-camp %n af?M h.s fi.) 'f ,1j/.} We-v$ r 72bw,s,.-. ~ m-u e ,emm, a 'p' h, t .s.t. t 1 i 1 o -ps,#' a., . m u{.-e c,c.,i :v 3 + i.2 o- ? p.3 a >N+3 er~ ?! 7 1,bn,,,s.[UNl4[;l 30$.-
- 7
' ? ? NDE-~/7Dd;'_9 w,3?;N,, d ' y,J 3 D" { I g'.g"'h@. ^ ^ 4'f.ili.g.fi kf '- d :' , y. ~pSji.?'7 ' ? r '4 2% *- ?. Q:g g$". f<. ^ %fy '3 m s l, Q~ Q ^ ..'my n: 'r a g x Specimen AW8 Tested at 20&F gN Kf's ~rp+n
- m. e
.,n.:m., ~,m.m_m,.g e.,4l' ,,. Li w m.... h m -:? W ' ( m y$to,J.C w h.4pMC 9q u y x ~~~ k l UT ~ 9'*( y 9:p p.. - ' td <",y,n g% g ' x' g I x.. 9 E f- ) p. 1 1 + e s m - m., a ? a. a 4 >$~h. .w m d es;sua d. % Y.Q Y' m,. - W~1 1 J +' 2. Specimen AW9 Tested at 55&F Figure 5-14 Fractured Tensile Specimens from Parley Unit 1 Reactor Vessel Surveillance Weld Metal 5-31
l-110.00 100.00-90.00-80.00-70.00-g' 80.00-E 50.00-N 40.00-30.00-AL7 20.00-150 F 10.00-0.00 0.00 0.10 0.20 STRAIN, IN/IN l 100.00 90.00-j 80.00-70.00-to 60.00- <ti 50.00-CCg 40.00-30.00-20.00-AL8 l 10.00-225 F 0.00 0.00' O.10 0.20 STRAIN, IN/IN L Figure 5-15 Engineering Stress-Strain Curves for Plate B6919-1 Tensile Specimens AL7 and AL8 [ (Longitudinal Orientation) 5-32 b t-
w 5 I'. t i i 100.00 90.00-l 80.00-70.00-U) M 60.00-Vi 50.00-M H 40.00-W 30.00-AL9 20.00-550 F 10.00-0.00 0.00 0.04 0.06 0.12 0.16 0.20 STRAIN, IN/lN 't 1 i C e Figure 5-16 Engineering Stress-Strain Curve for Plate B6919-1 Tensile Specimen AL9 l (Longitudinal Orientation) .i 5 5-33 t e
n . l '100.00 N I f 80.00-80.00-70.00- _ex. 80,00_ I 50.00-w CC g 40.00-l 30.00-20.00-AT 7 10.00-200 F r O.00 0.00 0.04 0.08 0.12 0.16 0.20 STRAIN, IN/IN 100.00 90.00- -[ 80.00-70.00-e* 80.00-vi 50.00-cc H 40.00-e 30.00-AT 8 i 20.00-250 F 10.00_ 0.00 0.00 0.04 0.08 0.12 0.16 0.20 i STRAIN, IN/IN Figure 5-17 Engineering Stress-Strain Curves for Plate B6919-1 Tensile Specimens AT7 and AT8 (Transverse Orientation) 5-34 2
l 100.00 1 80.00-i i 80.00-70.00-0) M 60.00-g 30.00-I.. m 1& 40.00-(0 30.00-AT 9 20.00_ l 10.00-550 F 0.00 0.00 0.04 0.08 0.12 0.1E STRAIN, IN/IN l Figure 5-18 Engineering Stress-Strain Curve for Plate B69191 Tensile Specimen A19 (Transverse Orientation) ( 5-35 l
100.00' so 00-80.00-70.00-m 80.00-50.00-w ocg 40.00-30.00-AW 7 i-20.00-10.00-75 F O.oo 0.00 0.10 0.20 i STRAIN, IN/IN I 100.00 90.00-l 80.00-70.00- ._m* 80.00-N 50.00-g ccg 40.00-1 30.00-AW 8 20.00-10.00-200 F o.oo o.00 0.10 0.20 STRAIN, IN/IN Figure 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens AW7 and AW8. p 5-36 t i
D77T. J' J e ', j i , bi 4: L i t .i i 100.00 i 90.00-80.00-e 70.00-(O M i 80.00-(O r 50.00-
- LLJ W
H 40.00-CO. 30.00-20. AW 9 i 10.00-550 F l 0.00 0.00-0.10-0.20 STRAIN, IN/IN l i r ? F ? Figure 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen AW9. f f t 5-37 I k i
'z: . p' I d' w f.) g, giO I cv -_ 3..e 6 s h L t i J f ..: i .( .( ~. t I I "
- 9
'b h ~ f , I . a ~ ? 9 4 6 j e s r 'p . 6 I - f h ..., +' 'y E .7 b h 1 9 '4 i .I t ^f I . 1 ~ I 1 F ? .} . f 'h I 6 I . I 4 m i -6
- l i
a ,t: .I r -l f 1 1 1 1 .5 + .i d 7 - f t. -: - -l
e i in f SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY . 6.1 Introduction l Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule r . geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two j reasons. First, in order to interpret the neutron radiation induced material propeety changes observed
- in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test f
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 environtnent at various positions within the pressure vessel and that experienced- ] by the test specimens. De former requirement is normally rnet by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. He latter information is generally derived solely from analysis. .i ne use of fast neutron fluence (E > 1.0 MeV) to conelate measured material property changes to the neutron exposure of the material has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent j years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance 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 en energy dependent damage j function for data correlation, ASTM Standard Practice E853, " Standard I'ractice for Analysis and j Interpretation of Light-Water Reactor Pressure Vessel Standard Surveillance Results", recommends l reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base - for future reference. De energy dependent dpa function to be used for this evaluation is specified in f t ASTM Standard Practice E693, " Standard Practice for Cha-cterizing Neutron Exposures in Ferritic 1 Steels in Terms of Displacements per Atom (dpa)". The application of the dpa parameter to the 3-assessment of embrittlement gradients through the thickness of the pressure vessel wall has already i 1 6-1 ) I j
been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials." This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule W, withdrawn at the end of the twelfth fuel cycle. Also included is an updated evaluation of the dosimetry contained in capsules Y, U, and X, withdrawn at the conclusion of cycles one, four and seven, respectively. This update is based on i current state-of-the-an methodology and nuclear data and, together with the capsule W results, provides a consistent up to date data base for use in evaluating the material propenies of the Farley Unit I reactor vessel. In each of the dosimetry evaluations, fast neutron exposure parameters in terms of neutron fluence (E > 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, uncenainties 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 neutron pad are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 107*,110,287*,290*,340*, and 343* 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 five feet of the 12-foot high reactor core. The presence of the surveillance capsules and associated support structures has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pad and the reactor vessel. In order to determine the neutran environment at the test specimen location, the capsules themselves must be included in the analytical model. 6-2 i i I
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 conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters ($(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios; i.e., [dpa/sec]/[$(E > 1.0 MeV)}, within the pressure vessel geometry. The relative radial gradient informatiors 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, where T is the beltline thickness of the reactor vessel). The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, $(E > 1.0 MeV), at surveillance capsule positions and at several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The source importance functions generated from these adjoint analyses provided the basis for all absolute exposure calculations and comparison with measurement. These importance functions, when combined with fuel cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the i 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 i 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 assemblies increased. The absolute cycle specific data from the adjoint evaluations together with the relative neutron energy spectra and radial distribution information from the reference forward calculation provided the means to: 1-Evaluate neutron dosimetry obtained from surveillance capsules. 2-Extrapolate dosimetry results to key locations at the inner radius and through the thickness of 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. 6-3
'Ihe forward transport calculation for the reactor model, presented in Figures 4-1 and 6-1, was carried t out in R,0 geometry using the DOT two-dimensional discrete ordinates code " Land the SAILOR cross-l section library"51 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 expansion of the scattering cross-sections and the angular discretization was modeled with an 3 i Se order of angular quadrature. The core power distribution utilized in the reference forward transport calculation was derived from statistical studies oflong-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 g fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, the neutron source was increased by a 20 margin derived from the statistical evaluation of plant-to-plant and cycle-to-cycle - l t variations in peripheral power. Since it is unlikely that any single reactor would exhibit power levels on the core periphery at the nominal + 20 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 P cross-3 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 run in R,0 geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case $(E > 1.0 MeV). .j Having the importance functions and appropriate core source distributions, the response of interest could be calculated as: R(r,0) = f f f f(r,0,E) S(r,0,E) r dr de dE reE where: R(r,0) = $(E > 1.0 MeV) at radius r and azimuthal angle 0. 1(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 $(E > 1.0 MeV), prior calculations"'I have shown that, while the 6-4
implementation of low leakage loading patterns significantly impacts both the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of [dpa/sec]/[$(E > 1.0 MeV)] is insensitive to changing core source distributions. In the application of these adjoint importance functions to the Farley Unit I 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 $(E > 1.0 MeV) solutions from the individual adjoint evaluations. The reactor core power distributions used in the plant specific adjoint calculations were taken from the fuel cycle design reports for the first twelve operating cycles of Farley Unit 1 """* *. I Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provides the means to correlate dosimetry results with the corresponding exposure of the pressure vessel wall. In Table 6-1, the calculated exposure parameters [$(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec] are 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 transport analysis, is meant to establish the absolute comparison of measurement with analysis. The reference data derived from the forward calculation is provided as a conservative exposure evaluation against which i plant-specific fluence calculations can be compared. Similar data are given in Table 6-2 for the pressur-vessel inner radius. Again, the three pertinent exposure parameters are listed for the reference and the cycle one through twelve plant-specific power distributions. It is important to note that the data for the vessel inner radius was taken at the clad / base metal interface, and thus represents the maximum predicted exposure levels of the vessel wall itself. Radial gradient information applicable to $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/see is given in Tables 6-3,6-4, and 6-5, respectively. The data, obtained from the reference forward neutron transport calculation, is 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 through 6-5. 6-5
y: y ?, ^ For example,'the neutron. flux $(E > 1.0 MeV) at the.1/4T depth in the pressure vessel wall along the : s O' azimuth is given by:- + $3j47(O = $(199.95, O F(204.95, O where: $Wf0')t Projected neutron flux at the 1/4T position on the 0* azimuth.- = . _ $(199.95,0*) Projected or calculated neutron flux at the vessel inner radius on the 0* =. azimuth. F(204.95,0') Ratio of the neutron flux at the 1/4T position to the flux at the vessel = i k inner radius for the 0* azimuth.' 'Ihis data is determmed from Table 6-3. ~o Similar expressions apply for exposure parameters expressed in terms of $(E > 0.1 MeV) and dpa/sec where the attenuation function F is obtained from Tables 6-4 and 6-5, respectively. 6.3 Neutron Dosimetsv The passive neutron sensors included in the Farley Unit I surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the - subsequent determination of the various exposure parameters of interest [$(E > 1.0 MeV), $(E > 0.1 - L MeV), dpa/sec]. The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. 'Ihe iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes - drilled in spacers at several axial levels within the capsules. The cadmium shielded uranium and L 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 - matenal 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 measurernents only if the irradiation parameters are well known. In particular, the following variables are of interest: The measured specific activity of each monitor. 'Ihe physical characteristics of each monitor. 6-6 m w...
i The operating history of the reactor.~ . 'Ihe energy response of each monitor. l 'Ihe neutron energy spectmm at the monitor location. f i The specific activity of each of the neutron monitors was determined using established ASTM i procedures * *""* *. Following sample preparation and weighing, the activity of each monitor was .i determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation. .l history of the Farley Unit I reactor during cycles one through twelve was supplied by NUREG-0020, " Licensed Operating Reactors Status Summary Report," for the applicable period. The irradiation f i history applicable to capsules Y, U, X and W is given in Table 6-7. I i i Having the measured specific activities, the physical characteristics of the sensors and the operating .j history of the reactor reaction rates referenced to full power operation were determined from the j i following equation: A R-P N FYE p C [1-e */] [e *'] i o j T I where R Reaction rate averaged over the irradiation period and referenced to operation at a core = t power level of P,,, (rps/ nucleus). Measured specific activity (dps/gm). l A = No Number of target element atoms per gram of sensor. l = Weight fraction of the target isotope in the sensor material. -F = Number of product atoms produced per reaction. Y = Average core power level during irradiation period j (MW) P = 3 P,,, = Maximum or reference power level of the reactor (MW). Calculated ratio of $(E > 1.0 MeV) during irradiation period j to the time weighted C = 3 average $(E > 1.0 MeV) over the entire irradiation period. A Decay constant of the product isotope (1/sec). =- length of irradiation period j (sec). 1 = 3 Decay time following irradiation period j (sec). t, = and the summation is carried out over the total number of monthly intervals comprising the irradiation period. 6-7
-~ ~ 1 .l In the equation describing the reaction rate calculation, the ratio [P]/[P,,,] accounts for month-by- - 3 - month variation of reactor cose power level within any given fuel cycle as well as over multiple fuel j cycles. The ratio C, which can be calculated for each fuel cycle using the adjoint transport technology. f 3 Ldiscussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux - h i level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single cycle irradiation, C is normally taken to be 1.0. However, for multiple cycle irradiations, particularly 3 . those employing low leakage fuel management, the additional C term should be employed. JIhe f 3 ' impact of changing flux levels for constant power operation can be'quite significant for sensor sets that [ have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low - leakage fuel management or for sensor sets contamed in surveillance capsules that have been moved from one capsule location to another. i c ~ For the irradiation history of capsules Y, U, X and W, 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 traction rates are listed { in Tables 6-8 through 6-11 for Capsules Y, U, X and W, respectively. i Values of key fast neutron exposure parameters were derived from the measured reaction rates using j the FERRET least squares adjustment code ". 'Ihe FERRET approach used the measured reaction rate j l data, sensor reaction cross-sections, and a calculated trial spectrum as input and pmceeded 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 then obtained from the adjusted spectrum. 3 i 1 In the FERRET evaluations, a log-normal least squares algorithm weights both the prior values and the - l 1 measured data in accordance with the assigned uncertainties and correlations. In general, the measured i values f are linearly related to the flux $ by some response matrix A: l ff) = E AE $7' E i where i indexes the measured values belonging to a single data set s, g designates the energy group, { i and cx delineates spectra that may be simultaneously adjusted. For example, b " b 'ig kg + i 6-8 i I
~. ._~ J 1 } l relates a set of measured reaction rates R, to a single spectrum $, by the multigroup reaction cross-section o The log-normal approach automatically accounts for the physical constraint of positive g t fluxes, even with large assigned uncertainties, j \\ 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. De trial input spectnam was converted to the FERRET 53 group structure using the SAND-II codc"1. His procedure was carried ' I out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a 1 SPLINE interpolation procedure in regions where group boundanes do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET. De sensor set reaction cross-sections, obtamed from the ENDF/B-V 4-2 =y file, were also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial a spectrum, as expanded to 620 groups, was employed as a weighting function in the cross-section. l i collapsing procedure. Reaction cross-section uncertamties in the form of a 53 x 53 covariance matnx for each sensor reaction were also constructed from the information contained on the ENDF/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 uncertamty information does not significantly impact the .l results of the adjustment. i i 1 Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to 3 the actual spectmm at the sensor set locations, the neutron spectmm input to the FERRET evaluation was taken from the center of the surveillance capsule modeled in the reference forward transport .j calculation. While the 53 x 53 group covariance matnces 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: l M,, = R,*, + R, R,1 P,1 l where R, specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set I of values. De fractional uncertainties R, specify additional random uncertainties for group g that are correlated with a correlation matrix given by: l i P,, = [1-0) 6,, + e e ' j 4 6-9 l
O i i where: i H= W 2 .2 y l 1 The first term in the conelation matrix equation specifies purely random uncertainties, while the .f r second term describes short range conelations over a group range.y(0 specifies the strength of the j latter term). 'Ihe value of 6 is I when g = g' and 0 otherwise. For the trial spectrum used in the ( t cunent evaluations, a short range correlation of y = 6 groups was used. This choice implies that. j neighboring groups are strongly correlated when 0 is close to 1. Strong long range correlations (or j W. Maerker's results anti-correlations) were justified based on information presented by R. E. Maerker t are closely duplicated when y = 6. e i The uncertainties associated with the measured reaction rates included both statistical (counting)'and f systematic components. The systematic component of the overall uncertainty accounts for counter t efficiency, counter calibrations, irradiation history corrections, and corrections for competag reactions in the individual sensors. l Results of the FERRET evaluations of the Capsules Y, U, X and W dosimetry are given in Tables - 6-12. The data summarized in the table includes fast neutron exposure evaluations in terms of 4(E > [ 1.0 MeV), @(E > 0.1 MeV), and dpa. Summaries of the fit of the adjusted spectrum to the measurements for each capsule are provided in Table 6-13. In general, good results were achieved in l l the fits of the adjusted spectra to the individual measured reaction rates. 'Ihe adjusted spectra from the least-squares evaluations are given in Tables 6-14 through 6-17 in the FERRET 53 energy group f structure. 1 A summary of the measured and calculated neutron exposure of Capsules Y, U, X and W is presented -l 1 in Table 6-18. The data listed in Table 6-18 indicates that, for the fast neutron evaluations, the ratios i of calculated to measured results were in the range of 0.911 to 1.261. 'Ihe results for Capsule W are j consistent with results obtained from similar evaluations of dosimetry from other Westinghouse .l reactors. f 3 I 6-10 i -. ~
6.4 Projections of Pressure Vessel Exposure l Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-19. Along with the current (12.43 EFPY) exposure, projections are also provided for exposure periods of 16,32, and 48 EFPY. In computing these vessel exposures, the calculated values from Table 6-2 were scaled by the average measurement / calculation ratios (M/C) observed from the evaluations of dosimetry from capsules Y, U, X and W for each fast neutron exposure parameter. This procedure resulted in bias factors of 1.006,0.909 and 0.936 being applied to the calculated values of 4(E > 1.0 MeV), $(E > 0.1 MeV), and dpa, respectively. Projections for future operation were based on the assumption that the average exposure rates characteristic of the cycle eight through twelve irradiation would continue to be applicable throughout plant life. The overall uncertan y associated with the best estimate exposure projections at the pressure vessel [ wall depends on the indiv;iual uncertainties in the measurement process, the uncertainty in the P dosimetry location, and on th uncertainty in the extrapolation of results from the measurement points to the point of interest in the ussel wall. For Farley Unit 1, an extrapolation uncertainty of 5% has been combined with a 10% unc ettainty in the plant-specific measurement / calculation bias factor derived from the four surveillance capsules to produce a net uncertainty of 11% in the projected i exposure of the pressure vessel wall. This 11% uncertainty applies at the la level for @(E > 1.0 MeV). i i Data based on both a 4(E > 1.0 MeV) slope and a plant-specific dpa slope through 'the vessel wall are { provided in Table 6-20 for exposure projections to 16,32, and 48 EFPY. i t In order to access RTuor versus fluence curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations: l WW4D =&(07)ECW4D E dra(OD i and K3'4D = +(07) @dpa(on aw4n 6-11 l 1
~ ] [? .Using this approach results in the dpa equivalent fluence values are listed in Table'6-19. In Table 6 '. l ~ i; 21' updated lead factors are listed for each of the Farley Unit I surveillance ciapsules. Lead factor data,- based on the accumulated fluence through cycle twelve, is provided for each remaining capsule. o I. s i j i .i i 6-12
d-FIGURE 6-1 PLAN VIEW OF A DUAL REACTOR VESSEL SURVEILLANCE CAPSULE - 16.94 DEG. -19.72 DEG. T T r r 73.31 IN. t +
- y sxxxx_xxy y A n k _' M A - Ax 6-13 gvme-
P x x t l ,s '$g-u ,TABG 61; ",,E ~ CALCULATED FAST NEUTRON EXPOSURE RATES AT SURVEILLANCE CAPSULE CENTERf . CAPSULE LOCATION - '17.0* 20.0* 2 WE >1.0 MeV). n/cm -sec CYCLE 1 . 1.603E+11 1.388E+11 - CYCLE 2'. ' 1.708E+11 -
- 1.470E+11 -
J CYCLE 3 ~ . 1.616E+11 - 1.399E+11. CYCG 4 1.877E+11 - 1.557E+11 - l CYCG 5 1.377E+11 1.201E+11 CYCG 6 1.238E+11 1.080E+11 - CYCG 7 1.345E+11 1.177E+11-CYCG 8 1.120E+11 : 9.827E+10 CYCLE 9 ' 1.234E+11 1.115E+11 . CYCG 10 1.366E+11 1.296E+11 - l . CYCG 11 - 1.095E+11 9.692E+10 - CYCLE 12 - 1.063E+11 9.095E+10 ' CRSD Data' 2.047E+11 1.727E+11 2 ME > 0.1 MeV). n/cm,,,, .3 CYCLE 1 8.497E+11 7.174E+11 - CYCG i 9.054E+11 7.597E+11 l CYCG 3 8.567E+11 .. 8.049E+11 7.232E+11 : CYCLE 4 9.950E+11 CYCLE 5 7.297E+11 6.207E+11 . CYCLE 6 6.562E+11 ' 5.581E+11 CYCG 7 7.128E+11 6.084E+11 CYCG 8 5.937E+11 5.081E+11 ' r CYCG 9 6.542E+11' 5.765E+11 CYCG 10 7.239E+11. 6.700E+11 CYCLE 11 5.802E+11 5.011E+11 ~ j CYCLE 12 5.633E+11 4.702E+11 CRSD Data 1.085E+12 8.927E+11 = 1ron Dielrenwat Rate. doa/sec CYCG 1 3.463E-10 2.955E-10 CYCLE 2 3.690E-10 3.130E-10 CYCLE 3 3.492E-10 2.979E-10 l CYCG 4 4.055E-10 3.316E-10 i CYCG 5 2.974E-10 2.557E-10 - i CYCLE 6 2.674E-10 ~ 2.299E CYCLE 7.. 2.905E-10 2.507E-10 CYCLE 8 2.419E-10 2.093E-10 CYCG 9 2.666E-10 2.375E-10 CYCG 10 2.950E-10 2.760E-10 CYCLE 11. 2.365E-10 2.064E-10 CYCLE 12 2.2%E-10 1.937E-10 CRSD Data' 4.422E-10 3.678E-10 ' l 6-14
Q h TABLE 6-2 e CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSUP.E RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE i N r 2 ME > 1.0 MeVL ' n/cm -sec O. 0* .]19* 20 1* ' 30.0*-- '45.0* ' CYCM 1 5.074E+10 3348E+10 2.818E+10 - ' 2.211E+10 1.572E+10 CYCLE 2 5.407E+10 3.580E+10 2.981E+10 2335E+10 : 1.649E+10 - CYCLE 3 5.023E+10 '. 3354E+10 2.838E+10 2.217E+10 1.574E+10f CYCLE 4 5.855E+10 4.007E+101 3.141E+10 2323E+10 1.573E+10 CYCG 5 4330E+10 2.824E+10 - 2.441E+10 1.810E+10 1.190E+10 CYCLE 6 2.811E410 2.464E+10 2.201E+10 1.695E+10 1.219E+10 - l CYCLE 7 ~ 4.183E+10 2.747E+10 ' 2397E+10 - 1.856E+10 -1332E+10 j CYCG 8 ' 3.108E+10. 2.240E+10 2.019E+10 -.1.583E+10 1.147E+10. CYCLE 9 3.257E+10 2375E+10 2.269E+10 1.737E+10 1.145E+10 - CYCM 10 3.518E+10 2.630E+10 2.664E+10 - 2.149E+10 1.154E+10 - CYCM 11 3.133E+10 2.190E+10 -
- 1.998E+10 1.600E+10 1.203E+10 CYCLE 12 3.157E+10 2.164E+10 1.868E+10 1.406E+10
- 1.057E+10 CRSD Data 6313E+10 4.256E+10 3.618E+10 2.789E+10 .1.984E+10 I i i 2 ME > 0.1 MeVL n/cm -sec
- 0. 0*
12.0* 20.5* - 3QO* 45.0* - Q* CYCLE 1 1.279E+11 8.237E+10 5.973E+10 4.599E+10 . 3.238E+10 CYCLE 2 1363E+11 8.806E+10 6319E+10 4.857E+10 3397E+10 CYCLE 3 1.266E+11 8.251E+10 6.017E+10 4.612E+10 3.243E+10 CYCLE 4 1.475E+11 9.856E+10 ' 6.660E+10 4.831E+10 3.240E+10 CYCLE 5 1.091E+11 6.947E+10 5.175E+10-3.764E+10 2.451E+10 CYCLE 6 7.085E+10 6.063E+10 4.666E+10 3.526E+10 2.511E+10 CYCLE 7 1.054E+11 6.759E+10 5.083E+10 3.861E+10 2.744E+10_ j CYCLE 8 7.833E+10 5.511E+10 4.280E+10 3.292E+10 - 2362E+10 j CYCLE 9 8.208E+10 5.842E+10 4.810E+10 3.612E+10 - 2358E+10 CYCLE 10 8.865E+10 6.469E+10 5.647E+10 4.470E+10 2377E+10 ' ' CYCLE 11 7.896E+10 5388E+10 4.235E+10
- 1.328E+10 2.479E+10 CYCG 12 7.956E+10 5323E+10.
3.961E+10 2.924E+10. 2.178E+10 ' CRSD Data - 1.591E+11 1.047E+11 - 7.671E+10 5.802E+10 4.087E+10 1 h 6-15 l 1
f{g }' l -I TABLE 6-2 ~ (CONTINUED)_ l CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES ' AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE i i t ~ i Ir -l Iron Atom Displacement Rate, doa/see i
- 0. O' 12.0*
20.5* 30.0* 45.0* CYCLE 1 8.119E-11 535E-11 4367E 3.427E-11 2.452E-11 i CYCLE 2 8.652E-11 : 5.728E-11 4.620E-11 3.620E-11 2.573E > CYCLE 3 8.037E-11 5366E-11 4399E 3.437E-11 2.456E-11 l CYCLE 4 9368E-11 6.410E-11 4.869E-11 3.600E-11 2.454E-11 i CYCLES 6.928E-11 4.518E-11 3.783E-11 2.805E-11 1.856E-11 ' CYCLE 6 4.498E-11 3.943E-11 3.411E-11 2.628E-11 1.901E-11 CYCLE 7 6.693E-11 43%E-11 3.716E-11 2.877E-11 2.078E-11 i CYCLE 8 4.973E-11 3.585E-11 3.129E-11 2.453E-11 1.789E-11 CYCLE 9 5.212E-11 3.800E-11 3.517E-11 2.692E-11 1.786E-11 ' CYCLE 10 5.629E-11 4.208E-11 4.129E-11 3331E-11 1.800E-11 CYCLE 11 5.013E-11 3.304E-11 3.097E-11 2.480E-11 1.877E-11 CYCLE 12 5.052E-11 3.462E-11 2.8966-11 2.179E-11 1.649E-11 CRSD Data 1.010E-10 6.809E-11 5.609E-11 4323E-11 3.095E-11 i P 1 I i i h i i 6-16 i i i
p.= 'i ,1, t TABLE 6-3 i -l Q-RELATIVE RADIAL DISTRIBUTION OF $ (E > 1.0 MeV) 1 WITHIN THE PRESSURE VESSEL WALL ~ i 1 i Radius i ifEll
- 0. 0* '
J.2.9*,- 20.j' ' E . 45.0* 199.955 1.00 1.00 1.00 1.00-1.00-200.91-0.926 0.924 0.926 ~0.927' O.927. 20230-0.799 0.796 0.801' O.801-0.803 i 203.74 .0.667 0.670 0.672 0.673 0.675, 1 205.13 0.554 0.560 0.559 0.562 0.564' 206.52 0.456 0.466 0.463 0.465 0.468 207.91 0374 0385 0381 0384 0387-l 20930 0306 0317 0312 0315 0318 210.69-0.249 0.260 0.255 0.258 - 0.261 212.07 0.202 0.212 0.209 0.211 0.214 j 213.46 0.164 0.173 0.170-0.172 0.175 i 214.85-0.132-0.140 0.138 0.140 0.143 216.24 0.105 0.113 0.111 0.113 0.116 { 217.63 0.0828 0.0904 0.0883' O.0900 0.0934 l 218.86 0.0650 0.0728 0.0708 0.0726 0.0766 i 219.95
- 0.0510 0.0586 0.0568
'O.0586 0.0628 ? NOTES:
- 1) Base Metal Inner Radius i
- 2) Base Metal Outer Radius
{ I 't 6-17 l l .\\ .n
g ,g TABLE 6-4 j RELATIVE RADIAL DISTRIBUTION OF $ (E > 0.1 MeV) [ - WITHIN 'INE PRESSURE VESSEL WALL l l Radius ' ,(em)_
- 0. O' 12.0*
20.,u * - 30.0* 410* 5 199.95m 1.00 1.00 1.00 1.00 1.00 200.91 1.00 1.00 1.00 1.00 1.00 f e 202.30 0.%5 0.970 0.978 0.978 0.984 e 203.74 0.905 0.917 0.926. 0.925-0.935 205.13 0.839 0.858 0.867 0.866 0.878 j 206.52 0.771 0.795 0.803 0.803 0.817 207.91 0.704 0.732 0.740 0.739 0.754 [ 209.30 0.639 0.668 0.676 0.676 0.693 f 210.69 0.575 0.606' O.614 0.614 0.632 212.07 0.513 0.546 0.554 0.554 0.573. 213.46 0.454 0.486 0.495 0.495 0.515 l 214.85 0.3% 0.429 0.438 0.438 0.458 216.24 0.339 0.372 0.382-0.382 0.403 217.63 0.283 0.316 0.327 0.327 0.350 218.86 0.230 0.265 0.277 0.279 0.303 219.95* 0.187 0.222 .2235 0.238 0.262 i I i NOTES:
- 1) Base Metal Inner Radius
- 2) Bese Metal Outer Radius f
6-18
[7: ( TABLE 6-5 RELATIVE RADIAL DISTRIBUTION OF dpa/sec l WITHIN THE PRESSURE VESSEL WALL -i i Radius j (cm)
- 0. 0*
12.0* 20.5* ' 30.0* 45.0" 199.95W. l l.00 1.00 1.00 1.00 1.00 l 2G9.91 0.939 0.939 0.938 0.938 0.938 20230 0.841 0.841 0.837 0.835 0.836 203.74 n739 0.744 0.732 0.731 0.732 205.13 0.648 0.657 0.640 0.639 0.641 l l 206.52 0.567 0.579 0.558 0.557 0.560 207.91 0.495 0.509 0.487 'O.486 -0.490 l 20930 0.432 0.447 0.425 0.423 0.428 210.69 0375 0391 0370 0369 0374 l 212.07 0325 0341 0321 0320 0326 213.46' O.280 0.2% 0.278 0.277. 0.283 . 214.85 0.239 0.255 0.240 0.239 . 0.245 l 216.24 0.202 0.217 0.204 0.204 0.211 i 217.63 0.166 0.182 0.172 0.171 0.180 218.86 0.134 0.152 0.145 0.145 0.155 I 219.955 0.108 0.127 0.122 0.123 0.133 ? NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius l
r ( t l r i l 6-19
~ ~- m t TABLE 6-6 l NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS' -[ t I Reaction. Target Fission l Monitor. of Weight ' Response Product Yield Material - Interest Fraction Ranae Half-Life ' (%) 'I 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 Ni"(n.p)Co" 0.6827 , E > 1.0 MeV 70.78 days f Uranium-233' U "(n,f)Cs'" 1.0 E > 0.4 MeV 30.17 yrs 6.00 2 l Neptunium-237' Np "(n,f)Cs'"' 1.0 E > 0.08 MeV 30.17 yrs - 6.27 2 Cobalt-Aluminum
- Co"(n,y)Co" 0.0015 0.4ev>E> 0.015 MeV 5.271 yrs Cobalt-Aluminum Co"(n 1)Co*
0.0015 E > 0.015 MeV 5.271 yrs .r
- Denotes that monitor is cadmium shielded.
l t t l I t 1 4 i i 5 t L 5
+, n
- TABLE 6 'If O
i MONTHLY THERMAL GENERATION DURING THE FIRST TWELVE FUEL CYCLES Thermal Generation 1hermal Generation - . OF 1NE FARLEY UNIT 1 REACTOR - ' Year - Month - (MW.hr) Year Month (MW.hr) - .g 1977-8 67,513 5 1,955,244.- f 9 864,971 6-1,875,991 10 355,140 7 1,968,060 t 11 1,131,938 8 1,679,390. 12 1,304,003 9: 1,812,488 1978
- 1..
1,354,896 10 1,710,147 i 2 1,426,824 11 1,909,130 3 1,884,641 12 1,856,957 4 1,662,759 1983 1 794,698 l 5 1,650,988 2 0-t 6 1,784,321 3 33,094 'r 7 1,830,416 4 1,489,953 8 1,797,336. 5 1,956,770 9 941,142 6 1,810,449 l 10 1,501,768 7 1,973,088. i l-11 1,896,861 8 1,971,688 I 12 1,787,293 9 1,909,440 1979 1 1,558,383 10 1,789,981 ~$ 2 1,616,052 11 .1,909,228 i 3 381,913 12 1,973,088- [ 4 0 1984 1 1,729,356 5 0 2-631,428 = 6 0 3 0 7 0 4 124,058 i 8 0 5' 1,807,741-l 9 0 6 1,908,735 10 0 7 1,973,062 11 457,481 8 1,973,088 l 12 1,760,937 9 1,904,651 } 1980 1 1,721,418 10 1,949,788 2-639,186 11 1,898,355 3 1,800,910 12 1,898,787 d 4 1,870,270 1985 1 1,973,088 t 5 1,927,148 2 1,782,144 6 822,411 3 1,708,471 7 1,201,139 4 - 329,500 l 8 1,694,628 5 85,286 ) 9 1,398,265 6 1,655,389 10 1,933,931 7 1,844,333 i 11 362,570 8' 1,958,242. 12 0 9. 1,871,312 1981 1 0 10 1,954,081 2 0 11 1,865,690 l 3 0 12 1,943,431 j 4 1,134,435 1986 1 1,877,709 5 1,639,348 2 1,741,224 6 1,836,579 3 1,812,234 7 1,889,655 4 1,872,495 8 1,902,225 5 1,853,814 9 552,776 6 1,888,407 i 10 0 7 1,846,629 l 11 0 8 1,775,600 12 0 9 1,890,035 1982 1 0 10 181,323 2 0 11 0 j 3 1,254,038 12 1,716,656 4 1,523,773 1987 1 1,751,070 I 6-21 l l
p - ? . TABLE 6-7 (CONTINUED) ~ 1987 2 1,768,852 7 1,971,762 3 1,955,044 8 1,971,823 4 1,216,358 9 1,564,342 5 1,906,563 10 0 6 1,909,366 11 257 7 1,973,035 12 1,560,307 8 1,748,771 1993 1 1,928,568 r 9 1,909,437 2 1,775,604 10 1,812,021 3 1,293,224 11 1,909,427 4 1,905,594 12 1,235,697 5 1,971,444 1988 1 1,973,088 6 1,908,087 2 1,845,755 7 1,972,236 3 1,581,648 8 1,972,212 4 0 9 1,907,885 5 388,449 10 1,975,209 6 1,773,947 11 1,909,034 7 1,973,088 12 1,972,228 8 1,970,025 1994 1 1,969,966 9 1,909,440 2 1,763,942 10 1,884,883 3 227,593 11 1,908,164 12 1,951,416 1989 1 1,964,305 2 1,757,472 3 1,932,709 4 1,892,334 5 1,973,088 6 1,906,099 7 1,967,625 8 1,969,301 9 1,383,665 10 0 11 855,994 12 1,968,102 1990 1 1,970,192 2 1,652,567 3 1,962,103 4 1,906,730 t 5 1,846,696 6 1,758,695 7 1,690,942 8 1,971,720 9 1,867,822 10 1,974,496 11 1,906,403 12 1,971,698 1991 1 1,970,245 2 1,780,012 3 488,986 4 0 5 467,378 6 1,777,930 7 1,919,659 8 1,725,229 9 1,890,966 10 1,879,448 11 1,909,440 12 1,961,350 1992 1 1,971,348 2 1,844,169 3 1,972,001 4 1,860,541 5 1,971,974 6 1,907,928 6-22
(n ,p ( A' r S' TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RA'IES g SURVFJIJ.ANCE CAPSULE Y SATURATED ACTIVITIES AND DERIVED FAST NEUTRON FLUX ~ MEASURED - ' SATURATED f REACHON ' MONTIM AND L ACTIVITY ' ACTIVITY ' RATE AXIAL LOCA710N (das/sec-am) (dis /sec-am) (rus/ nucleus) Cu-63 (nm) Co-60 ' 79-2712 TOP 6.590E+04 5.098E405 - 79-2717 MID 6.280E+04 4.858E+05 79-2722 BOT 6.810E+04 5.268E405~ Averages - 6.560E+04 5.075E+05 7.741 6 17 Fe-54 (n.0) Mn-54 79-2710 'IDP 1.950E+06 5.327E+06 79-2715 MID 1.900E+06 5.191E+06 79-2720 BOT 1.990E406 5.437E+06 Averages 1.947E+06 5.318E406 8.503 & l5 Ni-58 (n.0) Co-58 79-2711 TOP 1.010E+07 8.046E407. 79-2716 MID 9.630E406 7.671E+07 79-2721 BOT 1.040E+07 8.285E+07 Averages 1.004E+07 8.000E+07 1.142E-14 U-238 (n.f) Cs-137 (Cd) 79-2709 MID 2.590E+05 1.006E+07 6.628E-14 No-237 (a.f) Cs-137 (Cd) 79-2708 MID 1.820E+06 7.068E+07 4.437E-13 Co-59 (n.% Co-60 79-2713 TOP 1.440E+07 1.114E+08 79-2718 MID 1.500E+07 1.160E+08 79-2723 BOT 1.460E+07 1.129E+08 Averages 1.467E+07 1.135E+08 7.402E-12 Co-59 (n.% Co-60 (Cd) 79-2714 TOP 8.110E+06 6.274E+07 79-2719 MID 8.220E406 6.359E+07 79-2724 BOT 8.140E+06 6.297E+07 Averages 8.157E+06 6.310E+07 4.117E-12 I 6-23
c. TABLE 6-9 - MEASURED SENSOR ACTIVITIES AND REALTION R ATES f - SURVEILLANCE CAPSULE U. SATURATED ACITVITIES AND DERIVED FAST NEUTRON FLUX MEASURED SATURA7ED REACTION MONITOR AND ACTIVITY ACTIVTIY. RATE AXIAL LOCATION (dis /see-am) (dis /sec-am) (ros/ nucleus) 3 Cu43 (n.cx) Co-60 i Cu TOP 1350E+05 5.110E+05 Cu MID 1340E+05 5.072E+05 i-Cu BOT -1.420E+05 5375E+05 Averages 1370E+05 5.iB5E+05 7.910E-17 Fe-54 (n.o) Mn-54 Fe TOP 1.880E+06 5.557E+06 Fe MID 1.860E+06 5.498E406 Fe BOT 1.940E406 5.735E+06 Averages 1.893E+06 5.597E+06 8.949E-15 Ni-58 (n.o) Co-58 Ni TOP 4.160E406 8.685E+07 - Ni MID 4.030E+06 8.414E+07 Ni BOT 4330E+06 9.040E+07 Averages 4.173E+06 8.713E+07 1.244E 14 f U-238 (n.f) Cs-137 (Cd) U-238 MID 7.000E+05 1.068E+07 7.040E-14 No-237 (n f) Cs-137 (Cd) Np-237 MID 6.050E+06 9 233E+07 5.796E-13 [ Co-59 (n.M Co-60 f Co TOP 3.560E+07 1347E+08 Co MID 3.570E+07 1351E+08 Co BOT 3.600E+07 1363E+08 .f Averages 3.577E+07 1354E+08 8.832C-12 Co-59 (n.M Co-60 (Cd) Co 1.920E+07 7.267E+07 4 741E-12 e t 6-24 b
+ f P4 TABLE 6-10 i t MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE X ' SATURATED AC11VITIES AND DERIVED FAST NEUTRON FLUX f MEASURED SATURA7ED REACHON MONITOR AND - ACTIVITY - ACTIVITY -RATE AXIAL LOCATION (dis /see-am) (dis /see-am) (ros/ nucleus) [ Cu-63 (nn) Co-60 'i 87-1009 TOP 2.130E+05 4.738E+05 i 87-1014 MID 2.120E+05 4.715E+05 87-1019 BOT 2.220E+05. 4.938E+05 Averages 2.157E+05 4.797E+05 7.318E-17 Fe-54 (n.o) Mn-54 87-1011 TOP 2.270E+06 4.566E+06 87-1015 MID 2.240E+06 4.506E+06 -) 87-1020 BOT 2.400E+06 4.828E+06 Averages 2.303E+06 4.633E+06 7.408E-15 ~ -l Ni-58 (n.o) Co-58 l 87-1010 TOP 7.830E+06 7.243E+07 871016 MID ' 7.550E406 6.984E+07 j 87-1021 B OT 8.050E+06 7.447E+07 Averages 7.810E406 7.225E+07 1.032E-14 1 U-23R (n.f) Cs-137 (Cd) [ t 871005 MID 1.280E+06 1.018E+07 6.712E-14 l No-237 (n.f) Cs-137 (Cd) 87-1006 MID 9.880E+06 7.861E+07 4.935E-13 I Co-59 (n.% Co-60 87-1008 TOP 5.110E+07 1.137E+08 87-1013 MID 5.240E+07 1.166E+08 87-1018 BOT 4.760E+07 1.059E+08 Averages 5.037E+07 1.120E+08 7.309E-12 j Co-59 (n.M Co-60 (Cd) I 87-1007 TOP 2.840E+07 6.317E+07 ) 87-1017 BOT 2.640E+07 5.872E+07 Averages 2.740E+07 6.094E+07 3.976E-12 1 6-25 i i
y = - - - 1 .r ~, t m y a: .t . TABLE 6-11 .{ s - MEASURED SENSOR ACTIVII1ES AND REACTION RATES - SURVEILLANCE CAPSULE W ' ?!
- SATURATED ACITVITIES AND DERIVED FAST NEUTRON FLUX -
a 1 MEASURED SA'IURATED REACI1ON' d MONITOR AND-ACTIVITY ACTIVITY RA'IE l AXIAL LOCATION (dis /sec-ann (dis / soc-sm) fros/ nucleus) 1 Cu-63 (n.a) Co-60 i" 1621 TOP 2.360E+05 3.698E+05 94-1626 MID 2.350E+05 - 3.683E405.- 94-1631 BOT 2.490E+05 - 3.902E+05 Averages 2.400E+05 3.761E405 - 5.737E-17 j Fe-54 (n.o) Mn-54 l 94-1619 TOP 1.730E406 - 3.417E+06 1 94-1624 MID 1.700E+06-3.358E+06 l L. 94-1629 BOT 1.840E+06 . 3.634E+06 t Averages 1.757E+06 3.470E+06 5.548E-15 Ni-58 (n.o) Co-58 ' fl e 94-1620 TOP 8.460E+06 5.659E+07 ~ 94-1625 MID 8.080E+06 - 5.405E+07 Averages 8.443E+06 5.648E+07-8.064E-15 .[ 94-1630 BOT. 8.790E+06' 5.880E+07 - t U-238 (n.f) Cs-137 (Cd) l 94-1616 MID' 1.540E+06 6.519E+06 4.296E-14 i No-237 (n.f) Cs-137 (Cd)'- l 94-1615 MID ~ '1.150E+07 4.868E+07 3.056E-13 Co-59 (nN) Co-60 (Cd) I 94-1618 TOP 4.150E+07 6.503E+07 i[ 94-1623 MID 4.120E+07 6.456E+07 t 941628 BOT 4.150E+07 6.503E407 Averages 4.140E+07 6.488E+07 ' 4.233E-12 i l 6-26 i l
a -i . TABLE 6-12 <
SUMMARY
OF NEUTRON DOSIMETRY RESULTS - l SURVFHLANCE CAPSULES Y, U, X AND W
- j l
6 Calculation of Measured Huence for Capsule Y. . Flux .-Tune-Ruence Uncertainty Meas Fluence < 0.414 ev 1.267E+11 - 3.638E+07 4.609E+18 - 221 % - Meas Huence > 0.1 Mev 7.491E+11 3.638E407 '2.725E+19 15 % - Meas Fluence > 1.0 Mev 1.653E+11 3.638E+07 ' 6.014E+18 ' *8% - 'dpa 3.210E-10. 3.638E+07 1.168E-02 211 % I Calculation of Measured Muence for Capsule U : Flux Tune Huence Uncertainty' Meas Fluence < 0.414 ev 1.567E+11 9.735E+07.; 1.525E+19 220 % Meas Fluence > 0.1 Mev 9.106E+11 9.735E407 8.865E+19 tl5% Meas Fluence > 1.0 Mev - 1.866E+11 9.735E+07 1.817E+19 -
- 8% -
dpa 3.786E 9.735E+07 3.686E-02_ ' 211%, i Calculation of Measured Fluence for Capsule X Flux Tune Fluence. Uncertainty Meas Fluence < 0.414 ev 1.282E+11 1.928E+08 2.472E+19 220 % ' Meas Fluence > 0.1 Mev 7.893E+11 1.928E408 1.522E+20
- 15%
-l Meas Fluence > 1.0 Mev 1.588E+11 1.928E+08 3.062E+19
- 8%
,3, dpa 3.261E-10 1.928E+08 6.287E-02 - *11% Calculation of Measured Fluence for Capsule W . Flux Tinne ' Fluence: ' Uncertainty j ' 78% l Meas Fluence < 0.414 ev 2.245E+10 - 3.923E+08 8.806E+18 - Mens Fluence > 0.1 Mev 4.832E+11.' 3.923E+08 1.895E+20' 216 % l Meas Fluence > 1.0 Mev 1.030E+11 3.923E+08 4.040E+19 '
- 8%
l dpa 2.051E-10 3.923E+08 8.045E-02
- 11%.
I .i I 'l 'l 6-27 I
MR ^ 79 qs - TABLE 6-13 n L COMPARISON OF MEASURED AND FERRET CALCULATED ~ ~ REACI1ON RATES AT THE SURVF.HLANCE CAPSULE CENTER - SURVEIL LANCE CAPSULES Y, U, AND X ' 3 l SURVEILLANCE CAPSUL.E Yi ADJUSTED. j t REACIlON MEASURED CAIEULA110N ,G&[ 1 . Cu-63 (n a) Co-60 ' 7.74E-17 ' 7.92E-17 l 1.02 : j Fe-54 (n.p) Mn-54.- 8.50E-15 8.41E 0.99 = j Ni-58 (n.p) Co-58. 1.17E 14 1.16E-14 0.99 i U-238 (n,f) Cs-137 (Cd) 5.45E-14 5.01E-14 ' O.92 - Np-237 (n,f) Cs-137 (Cd) 4.37E-13 4.86E-13 ~ 1.11 Co 59 (n,1) Co-60 7.40E-12 7.32E-12 0.99 Co-59 (n,1) Co-60 (Cd) 4.12E-12 4.14E-12 '1.00 SURVEIILANCE CAPSULE U: ADJUSTED REACTION MEASURED CAIfULATION .K g Cu-63 (n.a) Co-60 7.91E-17 ~ 8.12617 1.03 Fe-54 (n.p) Mn-54 8.95E-15 8.86E-15 0.99 Ni-58 (n.p) Co 1.27E-14 1.25E-14 0.98 U-238 (n,f) Cs-137 (Cd) 5.52E-14 ~ 5.43E-14 ' O.98 Np-237 (n,f) Cs-137 (Cd) 5.71E-13 ' 5.99E-13 ' l.05 - Co.59 (n,1) Co-60 8.83E-12 8.73E-12 0.99 Co-59 (n.1) Co 60 (Cd) 4.74E-12 4.76E-12. 1.00-SURVEILLANCE CAPSULE X: ' ADJUSTED REACTION MEASURED CALCULATION ' G&[,, . Cu-63 (n,u) Co 7.32E-17 7.41E-17 1.01 Fe-54 (n.p) Mn-54 7.41E-15 ./E-15 1.01' Ni-58 (n p) Co-58 1.06E-14 1.0$E-14 " 0.99 U-238 (n.f) Cs-137 (Cd) 5.01E-14 4.65E-14 0.93 Np-237 (n,f) Cs-137 (Cd) 4.86E-13 5.12E-13 1.05 Co-59 (n,y) Co-60 7.31E-12 7.23E-12 0.99 Co 59 (n,1) Co-60 (Cd) 3.98E-12 4.00E-12 1.00 (: 6-28
TABLE 6-13 (CONTINUED). COMPARISON OF MEASURED AND FERRET CALCULA'IED REACTION RATES AT THE SURVEILLANCE CAPSULE CEN'IER i SURVEILLANCE CAPSULES W. e SURVEI11ANCE CAPSULE W: ADJUSTED -' REACTION MEASURED CALCULATION 'C/M Cu-63 (n,a) Co-60 5.74E-17 5.84E-17 1.02 Fe-54 (n p) Mn-54 5.55E-15 5.55E-15 1.00 Ni-58 (n.p) Co-58 8.06E-15 7.87E-15 0.98 6 U-238 (n.f) Cs-137 (Cd) 3.1IE-14 3.11E-14 1.00 Np-237 (n f) Cs-137 (Cd) 3.06E-13 3.22E-13 1.06 Co-59 (n,y) Co-60 (Cd) 4.23E-12 4.21E-12 0.99 t 1 9 3 6-29 .i
t .d. TABLE 6-14 i s
- ADJUSTED NEUTRON ENERGY SPECTRUM AT THE j
CENTER OF SURVEILLANCE CAPSUIE Y g ENERGY ~ ADJUSTED FLUX ENERGY ADJUSTED FLUX. -l . GROUP (MeV) (n/cm -sec) GROUP (MeV)' (n/cm*-sec) 2 1 1.733E+01 -. 1.038E+07 -29 5.531E-03 3.472E+10 2-1.492E401 2.370E+07 .30 3355E-03 1.158E+10 3-1350E+01 9316E+07 31 2.839E-03 1.160E+10 4 1.162E+01 2.123E+08 '32 2.404E-03 1.172E+10, l 5 1.000E+01 4.773E+08 33 2.035E-03 3372E+10 6 8.607E+00 8363E408 34 1.234E-03 2.971E+10 7 7.408E+00 - 1.968E+09 35 7.485E-04 2.695E+10 8 6.065E+00 2.879E+09 36 4.540E-04 2.582E+10 ' 9 4.966E+00 6.283E+09 37 2.754E-04 2.730E+10. l i 10 3.679E+00 8.538E409 38 1.670E-04 3.206E+10 11 2.865E+00 1.790E+10 39 1.013E-04 3.043E+10 l 12 2.231E+00 2.501E+10 40 6.144E-05 2.958E+10 13 1.738E+00 3.546E+10 41 3.727E45 2.861E+10 l 14 1353E+00 4.049E+10 42 2.260E-05 2.730E+10 .j 15 1.108E+00 7354E+10 43 1371E45 2.596E+10 l 16 8.208E-01 8.517E+10 44 8315E 06 2398E+10 i 17 6393E-01 8.827E+10 45 5.043E46 2.105E+10 18 4.979E-01 6.765E+10 46 - 3.059E-06 1.855E+10. 19-3.877E-01 9.769E+10 47 1.855E-06 1.616E+10 20 3.020E 01 8.777E+10 48 1.125E-06 1.168E+10 l 21 1.832E-01 9.284E+10 49 6.826E-07 1.408E+10 ' 22 1.111E-01 7.620E+10 50 4.140E-07 1.825E+10 t 23 6.738E-02 4.817E+10 51 2.511E-07 1.922E+10 24 4.087E-02 2.526E+10 52 1.523E-07. 1.821E+10 j 25 2.554E-02 4.057E+10 53 9.237E-08 7.003E+10 ( 26 1.989E42 1.559E+10 l 27 1.503E-02 1.611E+10 4 28 9.119E43. 2.803E+10 l Note: Tabulated energy levels represent the upper energy in each group. t 6-30 i i I f
y E r I TABLE 6-15 i ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVETI I-ANCE CAPSULE U. i ENERGY ADJUSTED FLUX ENERGY ADJUSTED FLUX - GROUP (MeV) (n/cm'-see) GROUP (MeV) {p'em-sec) i r 1 1.733E+01 1.093E+07 29 5.531E-03 3.994E+10 2 1.492E+01 2.475E+07 30. 3355E-03 1325E+10 3 1350E+01 9.655E+07 31 2.839E-03 1323E+10 t 4 1.162E+01 2.185E+08 32 2.404E-03 1335E+10 5 1.000E+0! 4.891E+08 33 2.035E 3.844E+10 6 8.607E+00 8.564E408 34 1.234E-03 3391E+10 i 7 7.408E+00 2.021E409 35 7.485E-04 3.079E+10 - 8 6.065E+00 2.983E+09 36 4.540E-04 2.959E+10 9 4.966E+00 6.591E+09 37 2.754E-04 3.143E+10 10 3.679E+00 9.083E+09 38 1.670E-04 3.694E+10 11 2.865E+00 1.933E+10 39 1.013E-04 3.4%E+10 f 12 2.231E+00 2.747E+10 40 6.144E45 3.421E+10 13 1.738E+00 3.997E+10 41 3.727E-05 3.284E+10 14 1353E+00 4.729E+10 42 2.260E-05 3.116E+10 15 1.108E+00 8.827E+10 43 1371E-05 2.954E+10 [ 16 8.208E-01 1.042E+11 44 8315E-06 2.723E+10 17 6393E41 1.094E+11 45 5.043E-06' 2384E+10 f 18 4.979E-01 8.442E+10 46 3.059E-06 2.097E+10 19 3.877E-01 1.224E+11 47 1.855E-06 1.825E+10 20 3.020E-01 1.099E+11 48 1.125E 1318E+10 21 1.832E-01 1.157E+11 49 6.826E-07 1.604E+10 22 1.IllE-01 9.414E+10 50 4.140E-07 2.103E+10 f 23 6.738E-02 5.887E+10 51 2.511E-07 2.269E+10 24 4.087E42 3.050E+10 52 1.523E-07 2304E+10 l [ 25 2.554E-02 4.837E+10 53 9.237E-08 9.003E+10 26 1.989E-02 1.837E+10 27 1.503E-02 1.879E+10 28 9.119E-03 3.245E+10 Note: Tabulated energy levels represent the upper energy in each group. i 6-31 l i I
,+ a p .4 - i TABLE 6-16' ADJUSTED NEUTRON ENERGI SPECT. RUM AT THE CENTER OF SURVEILLANCE CAPSULE X-ENERGY ADJUSTED FLUX ENERGY ADJUSTED FLUX 2 GROUP' (MeV) (n/cm -sec) GROUP-(MeV) (dem'-sec) I 1 1.733E+0! ,1.023E+07 29 5.531E-03 3.449E+10 'f .2 1.492E+01 2307E+07 30 3355E-03 1.143E+10 - 3 1350E+01 8.941E+07 31 2.839E-03 1.140E+10 i 4 1.162E401 2.006E+08 32 2.404E-03 1.147E+10 5 1.000E411 4.440E408 =33 2.035E-03 3.297E+10 6 8.607E+00 7.633E+08 34 1.234E-03 2.901E+10 7 7.408E+00 1.768E+09 35-7,485E-04 2.627E+10 8 6.065E+00 2.551E+09 36 4.540E-04 2.516E+10 I 9 4.966E+00 . 5.523E+09 37 2.754E-04 2.665E+10 10 3.679E+00 7.580E+09 38 1.670E-04 3.092E+10 - 11 2.865E+00 1.620E+10 39 1.013E-04 2.%1E+10 [-t 12 2.231E+00. 2332E+10 40 6.144E-05 2.900E+10. j 13 1.738E+00 3.417E+10 41 3.727E-05 2.791E+10 { l4 1353E400 4.031E+10 42 2.260E-05 2.656E+10 f 15 1.108E+00 7.569E+10 43 1371E-05 2.524E+10 16 8.208E-01 8.991E+10. 44 8315E-06 2330E+10 17 - 6393E-01 9.498E+10 45 5.043E-06 ' 2.042E+10 = 18 4.979E-01 7364E+10 46 3.059E-06 1.798E+10 j 19 3.877E-01 1.071E+11 47 1.855E-06 1.565E+10 20 3.020E-01 9.629E+10 48 1.125E-06 1.131E+10 t 21 1.832E-01 1.013E+11 49 6.826E-07 1369E+10 l 22 1.111E-01 8.237E+10 50 4.140E 07 1.785E+10 23 6.738E-02 5.142E+10 51 2.511E-07. 1.899E+10 24 4.087E.02 2.658E+10 52 1.523E-07 1.916E+10 t 25 2.554E-02 4.207E+10 53 9.237E-08 7.224E+10 26 1.989E-02 1.594E+10 27 1.503E-02 1.627E+10 i '28 9.119E-03 2.806E+10 i .t Note: Tabulated energy levels represent the upper energy in each group. i 6-32 1 i
[ TABLE 6-17 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE l s e CENTER OF SURVEn J.ANCE CAPSULE W i . i h ENERGY, ADJUSTED FLUX ENERGY ADJUSTED FLUX ' 2 GROUP (MeV)- (n/em -sec) GROUP (MeV) (n/cm'-sec) I ' l' 1.733E+01 8.ll5E+06 29 5.531E-03 2325E+10 2 1.492E+01 1.846E+07 30 3355E-03 7.824E409 3 1350E+01 7.174E+07. 31 2.839E 7.970E+09 4-1.162E+01 1.611E+08 32 2.404E43 S.255E+09 5 1.000E401 3.563E+08 33 . 2.035E-03 2.455E+10 - 6 8.607E+00 6.085E+08 34 1.234E-03 2.260E+10 ,j 7 7.408E400 1396E409 35 7.485E-04 2.155E+10 8 6.065E+00 1.984E409 36 4.540E44 2.171E+10 9 4.966E+00 4.155E+09 37 2.754E-04 2386E+10 10 3.679E+00 5.426E+09 38 1.670E-04 . 3.483E+10 - 11 2.865E+00 1.114E+10 39 1.013E-04 2.626E+10 12 2.231E400 1.530E+10 40 . 6.144E 2.465E+10 13-1.738E400 2.157E+10 41 3.727E-05 2.241E+10 i 14 1353E+00 ' 2.508E+10 42 2.260E-05 2.00$E+10 l 15 1.108E+00 4.590E+10 43 1371E-05 1.801E+10 16 8.208E-01 5387E+10 44 8315E 06 ' l.582E+10 17 6393E-01 5.647E+10 45 5.043E-06 1323E+10 l 18 4.979E-01 4.416E+10 - 46 3.059E-06 1.119E+10 19' '3.877E-01 6.460E+10 47 1.855E-06 9.446E+09 ' 20 3.020E-01 5.832E+10 48 1.125E-06 6.665E+09 21 1.832E-01 6.184E+10 49 6.826E47 5.968E+09 - 22 1.111E-01 5.125E+10 50 4.140E-07 5.939E+09 1 23 6.738E-02 3.224E+10 51 2.511E-07 4.817E+09 ] 24 4.087E-02 1.691E+10 52 1.523E-07 3.922E+09 25 2.554E-02 2.694E+10 53 9.237E-08 7.771E+09 - .l 26 1.989E-02 1.031E+10 i 27 1.503E-02 1.080E+10 t 28 9.119E-03 1.858E+10 Note: Tabulated energy levels represent the upper energy in each group. t 6-33 1
g TABLE 6-18 COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FOR FARLEY UNIT 1 SURVEII.T.ANCE CAPSULES Y, U, X AND W Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule Y Calculated Measured C/M Fluence (E > 1.0 Mev). n/cm2 5.832E+18 6.014E+18 0.970 Fluence (E > 0.1 Mev), n/cm2 3.091E+19 - 2.725E+19 1.134 dpa 1.260E-02 1.168E-02 1.079 i Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule U Calculated Measured C/M Fluence (E > 1.0 Mev), n/cm2 1.656E+19 1.817E+19 0.911 Fluence (E > 0.1 Mev), n/cm2 8.777E+19 8.865E+19 0.990 dpa 3.577E-02 3.686E-02 0.970 Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule X - Calculated Measured - C/M Fluence (E > 1.0 Mev), n/cm2 2.917E+19 3.062E+19 0.953 Fluence (E > 0.1 Mev), n/cm2 1.546E+20 - 1.522E+20 1.016 dpa 6.301E-02 6.287E-02 1.002. Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule W Calculated Measured C/M Fluence (E > 1.0 Mev), n/cm2 4.621E+19 4.040E+19 1.144 Fluence (E > 0.1 Mev), n/cm2 2.389E+20 1.895E+20 1.261 dpa 9.843E42 8.045E42 1.223 k 7 s.
RM 4 c; TABLE 6-19 1 _.m< -i NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ~' ] ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFA'CE ' } a BEST ESTIMATE EXPOSURE (12.43 EFPY) AT THE PRESSURE VESSEL INNER RADIUS - .0DEG 12 DEG. 20.5 DEG l 30 DEG.. 45 DEG c. E > 1.0 1.539E+19 1.073E+19 - 9.486E+18 - 7.354E+18 '5.059E+18 .- E > 0.1 3.504E+19 2.386E+19. 1.817E+19 - 1.382E+19 9.417E+18 i dpa 2.289E-02 1.597E-02 1.367E-02 1.060E 7.338E-03
- i' BEST ESTIMA'IE EXTRAPOLATION FLUX AT THE PRESSURE VESSEL INNER RADIUS -
0DEG 12 DEO 20.5 DEG 30 DEG 45 DEG = E > 1.0 3.256E+10 - 2.336E+10 2.181E+10 ' 1.709E+10 1.149E+10 E > 0.1 7.413E+10 5.193E+10 - 4.176E+10 3.213E+10 2.139E+10 - -l dpa 4.843E-11 3.475E-11 3.142E-11 2.463E-11 1.667E-11 -l BEST ESTIMATE EXPOSURE (16.0 EFPY) AT THE PRESSURE VESSEL INNER RADIUS ~ 0DEG 12 DEG 20.5 DEG 30 DEG 45 DEG E > 1.0 1.906E+19 1.336E+19 1.194E+19 9.279E+18 ' 6.353E+18 - I E > 0.1 4.339E+19 2.971E+19 2.287E+19 .1.744E+19 1.183E+19. ~ j dpa 2.835E-02 1.988E-02 1.721E-02 1.337E-02 9.216E-03 ( I BEST ESTIMATE EXPOSURE (32.0 EFPY) AT THE PRESSURE VESSEL INNER RADIUS 0DEG 12 DEG. . 20.5 DEG 30 DEG 45 DEG j E > 1.0 3.550E+19 2.516E+19 2.296E+19 1.791E+19 1.216E+19 'f E > 0.1 8.082E+19 5.593E+19 4.3%E+19 3.366E+19 2.263E+19 dpa 5.280E-02 3.743E-02 3.307E-02 2.581E-02 1.763E-02 i BEST ESTIMATE EXPOSURE (48.0 EFPY) AT THE PRESSURE VESSEL INNER RADIUS 'i 0DEG 12 DEG 20.5 DEG 30 DEG 45 DEG E > 1.0 5.194E+19 3.695E+19 3.397E+19 2.654E+19 1.7%E+19 E > 0.1 1.183E+20 8.215E+19 6.505E+19 4.989E+19 3.343E+19 dpa 7.725E-02 5.498E-02 4.894E-02 3.825E-02 2.605E-02
- l 6-35
l
- c TABLE 6-20 f
NEUTRON EXPOSURE VALUES FOR THE FARLEY UNIT 1 l REACTOR VESSEL i FLUENCE BASED ON E > 1.0 MeV SLOPE a O DEG 12 DEG 20.5 DEG 30 DEG ' 45 DEG - 16 EFPY FLUENCE SURFACE 1.906E+19 1.336E+19 1.194E+19 9.279E+18 6.353E+18 1/4T. 1.085E+19 7.669E+18 6.854E+18 5.345E+18 ' 3.672E+18 I 3/4T 2.478E+18 1.844E+18 1.624E+18 1.281E+18 8.958E+17 32 EFPY FLUENCE-g SURFACE 3.550E+19 2.516E+19 2.2%E+19 1.791E+19 1.216E+19 1/4T 2.020E+19 1.444E+19 1.318E+19 1.032E+19 7.028E+18 3/4T 4.615E+18 3.472E+18 3.123E+18 2.472E+18 1.715E+18 48 EFPY FLUENCE SURFACE 5.194E+19 3.695E+19 3.397E+19 2.654E+19 1.7%E+19 1/4T 2.955E+19 2.121E+19 1.950E+19 1.529E+19 1.038E+18 3/4T 6.752E+18 5.099E+18 4.620E+18 3.663E+18 2.532E+18 FLUENCE BASED ON dpa SLOPE ODEG 12 DEG 20.5 DEG 30 DEG 45 DEG 16 EFPY FLUENCE ? SURFACE 1.906E+19 1.336E+19 - 1.194E+19 9.279E+18 6.353E+18 1/4T 1.258E+19 8.924E+18 7.785E+18 6.041E+18 4.149E+18 .i 3/4T 4.498E+18 3.367E+18 2.830E+18 2.190E+18 1.544E+18 32 EFPY FLUENCE SURFACE 3.550E+19 2.516E+19 2.2%E+19 1.791E+19 1.216E+19 1/4T 2.343E+19 1.681E+19 1.497E+19 1.166E+19 7.940E+18 t 3/4T 8.378E+18 6.340E+18 5.442E+18 4.227E+18 2.955E+18 48 EFPY FLUENCE SURFACE 5.194E+19 3.695E+19 3.397E+19 2.654E+19 1.7%E+19 1/4T 3.428E+19 2.468E+19 2.215E+19 1.728E+19 1.173E+19 3/4T 1.226E+19 9.311E+18 8.051E+18 6.263E+18 4.364E+18 [ t-6-36
am, f; ..f, . TABLE 6-21 UPDATED LEAD FACTORS FOR FARLEY UNIT 1. s, SURVEILLANCE CAPSULES l CAPSULE IEAD FACTOR WITHDRAWN Y-3.11 'EOC1 U 3.18 EOC4 X 3.30 - EOC7 l W* 3.02 EOCl2 V 3.02 Z 3.02
- BASIS FOR THIS ANALYSIS 1
h e 3 h -i 6 l i I i t i I 6-37 ) i i i i ~
SECTION 7.0 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE The following withdrawal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Farley Unit I reactor _ vessel: l TABLE 7-1 Surveillance Capsule Withdrawal Schedule Capsule Location Removal Fluence-Capsule (Degree) Lead Factor EFPY(*) (n/cm ) 2 Y *) 343 3.11 1.13 5.83 x 10 U *) 107 3.18 3.02 1.65 x 10 X *) 287 3.30 6.12 2.80 x 10 W *) 110 3.02 12.43 4.040 x 10 V 290 3.02 Standby 'Z 340 3.02 Standby NOTES: (a) Effective Pull Power Years (EFPY) fmm plant stanup. (b). Plant-specific evaluation. 7-1
x ..y
- . C h
g' SECTION 8.0 : j REFERENCES' 1
- f
~ d 1.. . Regulatory Guide 1.99, Revision 2, Radiation Embrinlement of Reactor Vessel Materials, U.S. j Nuclear Regulatory Commission, May 1988. I 1 2. Code of Federal Regulations,10CFR50, Appendix G. Fracture Toughness Requirements, U.S.- l-Nuclear Regulatory Commission, Washington, D.C. .[
- i 3.
WCAP-8810, Southern Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.'1 ...+ Reactor Vessel Radiation Surveillance Program,1. A. Davidson, et aL, December 1976. i 4. Section III of the ASME Boiler and Pressure Vessel Code, Appendix G. Protection Against'- 'l Nonductile Failure. ) I i 5. - ASTM E208, Standard Test Methodfor Conducting Drop-Weight Test :o Determine l Nil-Ductility Transition Temperature of Ferritic Steels, l 6. Southern Nuclear Operating Company letter to the U. S. Nuclear Regulatory Commission,. j . Joseph M. Farley Nuclear Plant Responses to Open issues Regarding Generic Exner 92-01, Revision 1, Reactor Vessel StructuralIntegrity,1. D. Woodard, dated June 21,1994. O I 7. Code of Federal Regulations,10CFR50, Appendix H. Reactor vessel Material Surveillance j Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C. j l 8. ASTM El85-82, Standard Practice for IJght-Water Cooled Nuclear Power Reactor Vessels, E706 (IF). .l l 1 9. ASTM E23-93a, Test Methodsfor Notched Bar impact Testing ofMetallic Materials. i 10. ASTM A370-92, Standard Test Methods and Depnitionsfor Mechanical Testing of Steel Products. l 1 11. ASTM E8-93, Test Methods of Tension Testing of Metallic Materials. 8-1 j I s
g.3 ,[ n 't n. e '12: .AS'1% E21-92, Standard Practicefor Elevated Temperature Tension Tests ofMetallic ^ Materials. l 13. ASTM E83-93, Practicefor Verification and Classification of Extensometers.' ,l '!1 14. ' WANL-PR(LL)-034, Vol. 5, Nuclear Rocket Shielding Methods, Mods)ication, Updating and ' 3 . Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique, 1 R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, August 1970. l f q. y y '15 ORNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded,' 47 Neutron, 20 . Gamma-Ray, P3, Cross Section L.ibraryfor Light Water Reactors. l t r l' 16. Nuclear Science and Engineering, Volume 94, Accountingfor Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis, R. E. Macrker, et al., Pages 291-308,. l 1986. q -i 17. WCAP-8515. The Nuclear Design of the Joseph M. Farley Unit 1 Power Plant, Cycle 1, R.' E.. Radcliffe, March,1975. l8. WCAP-9467, The Nuclear Design and Core Management of the Joseph M. Farley Unit ] Power Plant, Cycle 2. J. C. Miller, R. E. Radcliffe, January,1979. j 1 19. ND5-79-096, Revised Nuclear Design Data for Cycle 2 of Farley Unit 1, R. E. Radcliffe, R. l A. Holmes, April 25,1979. j
- i 20.
WCAP-9761, The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 l Power Plant, Cycle 3, U. L. Brown, M. M. Baker, R. E. Radcliffe, July,1980. 21. ,WCAP-10036, The Nuclear Design and Core Management of the Joseph M. Farley Unit 1. .i . Power Plant, Cycle 4A, U. L. Brown, G..A. Love, January,1982. j -i l I 8-2 -t
- I22, WCAP-10308, & Nuclear Design and Core Management of the' Joseph M. Farley Unit 1 Power Plant, Cycle 5, D. Y. Chung, B. A; Palmer, U. L. Brown, K. W. Bonadio, L. A. Klotz,
{ .e April,1983. i ~ 23. . WCAP.10525, The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant, Cycle 6, R. D. Erwin, D. Y. Chung, U. L. Brown, K. W. Bonadio, April,1984. 24. WCAP-10795, & Nuclear Design and Core Management of the Joseph M. Farley Unit 1 r y Power Plant, Cycle 7 R. D. Erwin, K. W. Bonadio, September,1985. p ~ 25. ' WCAP-11291, The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 l Power Plant, Cycle 8, R. D. Erwin, K. W. Bonadio, November,1986. I 26. 'WCAP-11755,' Rev.1, The Nuclear Design and Core Management of the Joseph M. Farley. Unit 1 Power Plant, Cycle 9, R. F. Schmidt, K. W. Bonadio, May,1988.' 27. WCAP-12371, The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant, Cycle 10. R. F. Schmidt, R. M. Smith, R. Y. Yeh, September,1989. 28. WCAP-12869, Rev.1, & Nuclear Design and Core Management of the Joseph M. Farley l Unit 1 Power Plant, Cycle 11, R. E. Radcliffe, K. W. Bonadio, April,1991. i WCAP-13434, The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 29. Power Plant, Cycle 12, R. E. Radcliffe, K. W. Bonadio, October,1992. l 30. WCAP-139C9, The Nuclear Design and Core Management of the Joseph M. Farley Unit 1, l Cycle 13 R. E. Radcliffe, J. Kastanes, K. W. Bonadio, April,1994. l 31. ASTM Designation E482-89, Standard Guidefor Application ofNeutron Transport Methods for Reactor VesselSurveillance,in ASTM Standards, Section 12, American Society for l Testing and Materials, Philadelphia, PA,1993. 8-3 i I ~l
~ d 9 i 32."
- ASTM Designation E560-84, Standard Recommended Practicefor Extrapolating Reactor ~
]
- Yessel Surveillance Dosimetry Results, in ASTM Standards, Section 12, American Society for6 -
Testing and Materials, Philadelphia, PA,1993. -) i .i 33. ASTM Designation E693-79, Standard Practicefor Characterizing Neutron Expos'ures in Ferritic Steels in Terms of Displacements per Atom (dpa), in ASTM Standards, Section 12,L American Society for Testing and Materials, Philadelphia, PA,1993. q 't - 34. ASTM Designation E706-87, Standard Master Matrixfor Light Water Reactor Pressure Vessel . Surveillance Standard, in ASTM Standards, Section 12, American Society for Testing and 0 ' Materials, Philadelphia, PA,1993. ] 35. ASTM Designation E853-87, Standard Practicefor Analysis and h*terpretation of-l i Light-Water Reactor Surveillance Results, in ASTM Standards, Section 12,' American Society: i for Testing and Materials, Philadelphia, PA,1993. 1 d ~ 36. 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,1993.- !;i ? 37. ASTM Designation E262-86, Standard Methodfor Measuring Thermal Neutron Flux by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing - 'l and Materials, Philadelphia, PA,1993. r 1 l 38. ASTM Designation E263-88, Standard Methodfor Determining Fast-Neutron Flut Density by ' Radioactivation ofIron, in ASTM Standards, Section 12, American Society for Testing and q Materials, Philadelphia, PA,1993. q 39. ASTM Designation E264-92, Standard Methodfor Determining Fast-Neutron Flux Density by l -l Radioactivation ofNickel, in ASTM Standards, Section 12, American Society.for Testing and ' Materials, Philadelphia, PA,1993. l 8-4 4 m. m + inv--
J ~ L40. J ' ASTM Designation E481-92, Standard Methodfor Measuring Neutron-Flux Density by. Radsoactivation of Cobalt and Silver, in AS'1M Standards, Section 12,' American Society for : N Testing and Materials, Philadelphia, PA,1993. 41.' ASTM Designation E523-87, Standard Methodfor Determining Fast-Neutron Flux Density by Radsoactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and '. Materials, Philadelphia, PA,1993. 42.' ASTM D$signation E704-90, Standard Methodfor Measuring Reaction Rates by Radioactivation of Uranium-238, in ASTM Standards, Section 12,1 American Society fori Testing and Materials, Philadelphia, PA,1993. . 43. ASTM Designation E705-90, Standard Methodfor Measuring Fast-Neutron Flux Density by. Radioactivation ofNeptunium 237,in ASTM Standards, Section 12, American Society for-Testing and Materials, Philadelphia, PA,1993. ~ 44.. ASTM Designation E1005-84, Standard Methodfor Application and Analysis of Radsometric Monitorsfor Reactor Vessel Surveillance, in ASTM Stmiards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993. ] i .45. HEDI TME 79-40, FERRET Data Analysis Core, F. A. Schmittroth, Hanford Engmeering Development Laboratory, Richland, WA, September 1979. 46. AFWIrTR-7-41, Vol. I-IV, A Computer-Automated iterative Method ofNeutron Flux Spectra ; Determined by Foil Activation, W. N. McElroy, S. Berg and T. Crocket, Air Force Weapons - 1 Laboratory, Kirkland AFB, NM, July 1967. 47, EPRI-NP-2188, Development and Demonstration of an Advanced Methodologyfor LWR Dosimetry Applications, R. E. Macrker, et al.,1981. 'j 1 4 8-5
a. I, i APPENDIX A LOAD-TBE RECORDS FOR CHARPY SPECBEN TESTS 6 1 I i I A4 i 1 i 4 -), W-
CuNs 784472-AF13 W r W g p -PM - Maximim Load P - Fast Fracture Load y p P GY " Yield Load 1 l P = Fast Fracture a p u i Arrest Load I l I l l 1 I I I I I I I I I I I I I I I i 1 1 I I I I I l
- hY tM t
p Time GY = me to NnedMng Wg - Fracture initiation Region t Wp = Fracture Propagation Region t = Time to Maximum Load M l t = Time to Fast (Brittle) Fracture Start j p Fig. A-1-idealized load-time record i I =u.__ m
Farly et
- W" AL41 BASE t.!Mb i
e i \\ i I.- e a 5 *- e 'I u-1 1 .3 1.2 2.4 3.6 4.8 6.0 TIE ( MEC > FARLEY #1 "W" SPECIMEN NUMBER
- AL41 MATERIAL
- BASE LONG CAPSULE
- FARLEY #1'
- "W" CAPSULE ramuty et "W" 4L40 Sa3E (DC g
Ie i 8 s l 5 *- n w -l I~ 9- .5 12 2.4 3.6 4.8
- 6. 0 TIE
( MIEC ) FARLEY #1 "W" l SP"CIMEN NUMBER
- AL40 l
MATERIAL
- BASE LONG l
CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-2. Load time records for Specimens AL41 and AL40 A-2 i
- p. -
i FNILEY S1 "W" K42 BASE LDC e 4 5 4 4 5 I 7-7 3 *- i n w l =_ G ____..____ .5 1.e e.4 s.s 4.s
- s. e i
TIE ( IgEC 3 FARLEY #1 "W" { SPECIMEN NUMBER
- AL42 MATERIAL
- BASE LONG r
CAPSULE
- FARLEY #1
- "W" CAPSULE t
rA aEv et *W* AL43 BASE UNC 1 a_ e I sv 1 f e l ~ i =_ i ~ i 1 l i .5 1.2 a.4 3.s 4.0 s.o TIE ( MSEC ) j FARLEY #1 "W" SPECIMEN NUMBER
- AL43 MATERIAL
- BASE LONG CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-3. Load-time records for Specimens AL42 and AIA3 A-3 i
FflN.EY e1 *W* mL2E gast t.gpc i g I I e- ~ 1 a 4 51-M i .n. - 9 1.2 2.4 4.8
- 6. 0
~, TIE ( MBEI: ) FARLEY #1 "W" SPECIMEN NUMBER
- AL32 MATERIAL
- BASE LONG CAPSULE
- FARLEY #1'
- "W" CAPSULE ranLEY of
- W*
AL34 BASE IRC i i i 4 I..- .I E a S 1-M w e-t b f n-i TIE ( MMI: ) FARLEY #1 "W" l SPECIMEN NUMBER
- AL34
[ MATERIAL
- BASE LONG CAPSULE
- FARLEY #1
- "W" CAPSULE l
Figure A-4. Load. time records for Specimens AL32 and AL34 A-4 L t
r .1 1 rARLEY 95 *W M 3P EME UES i i i i i e "e-I* -m A l S *- ~ 1 n l u-TIE (M> FARLEY #1 "W" SPECIMEN NUMBER
- AL37 l
MATERIAL
- BASE IDNG CAPSULE
- FARLEY #1
- "W" CAPSULE ram.tv e r an.at ener uns i
a g e-s A S *- n w e-u- .3 1.2 a.4 3.. 4.. .9 TIE (M) FARLEY #1 "W" SPECIMEN NUMBER
- AL31 MATERIAL
- BASE LONG CAPSULE
- FARLEY #1
- "w" CAPSULE Figure A-5. Load-time records for Specimens AL37 and AL31 A-5
i ) i I rasury en u-Ess anst tac i e a s i I =_ E 5 *n - v ~ [ B =_ I j l e .9 1.2 2.4 s.6 4.0 6.0 TIE ( MSEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AL33 MATERIAL
- BASE IDNG CAPSULE
- FARLEY #1
- "W" CAPSULE ramtry et 'u-m44 east in c I_e.
w 9 m v l 1" w_ l e .s s.: a.4 s.s 4.s s.o TIE ( MEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AL44 l
MATERIAL
- BASE LONG CAPSULE
- FARLEY #1 l
- "W" CAPSULE j
Figure A-6. Load-time records for Specimens AL33 and AL44 l 1 A-6 t J
s: i ? ) FMILIV 81 "W" AL39 BASC LSC 3 5 B 5 y l k* ~ z a 51-n w e- .o 1.a a.4 3.s 4.s 6.e TIE ( HKC ) FARLEY #1 "W" SPECIMEN NUMBER
- AL39 MATERIAL
- BASE LONG CAPSULE
- FARLEY #1
- "W" CAPSULE rAnu[Y 81 "W*
4.38 BASE UBC 5 "e_ Ie z a S 9-n w q.. t o - i .9 1.2 2.4 3.s 4.0 6.0 TIE ( MBEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AL38 MATERIAL
- BASE IANG CAPSULE
- FARLEY #1
- "w" CAPSULE Figure A-7. Load-time records for Specimens AL39 and AL38 A-7
(. U FRE EY St *W* MAS OflSC IRC g I9-5 t n. 4 a s 9-n w y- = .9 1.2 2.4 3.6 4.8
- 6. 0 -
TIE C MBEC > FARLEY #1 "W" SPECIMEN NUMBER
- AL45 MATERIAL
- BASE LONG.
CAPSULE
- FARLEY #1
- "W" CAPSULE
~ resuv en "W* AL36-BASE UDC g O g-5. E a s s-a w ~ ~ q- .5 1.2 2.4 3.6 4.8 6.0 TIE ( MBEC ) + FARLEY #1 "W" l SPECIMEN NUMBER
- AL36 l
MATERIAL
- BASE LONG l
CAPSULE
- FARLEY #1
- "W" CAPSULE 1
Figure A-8. Load-time records for Specimens AL45 and AL36 l l i l A-8 l-
i ) - ranutr en u-m.ss east unc I 3 5 4 y i ;~ s 31-e 1 l l i l I o .9 1.e a.4 3.6 4.s
- 6. o TIE C IWEC )
FARLEY #1 "W" SPECIMEN NUMBER
- AL35 MATERIAL
- BASE LONG CAPSULE
- PARLEY #1
- "W" CAPSULE i
rua.rv e u-at44 anse vanns i j "1 m _ 4 s A _f' 3 g_ n e Y ~ Y __ o .9 1.2 2.4 3.6 4.8
- 6. 0 t
TIE ( MIEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT44 MATERIAL
- BASE TRANS i
CAPSULE
- FARLEY #1-
- "W" CAPSULE i
Figure A-9. Load-time records for Specimens AL35 and AT44 A-9
4 'l rdW!MY St *W" 4745 sASE TRAMS l 4 I e_ "a 4 a s n-n I e i Y ( m-4 .5 1.2 E.4 3.6 4.0 6.0 TIE ( MEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT45 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE ranay si au-4Ty sAst u ns i
9 e_ 3, E 6 m 51-w N_ f 8 = 1.t' t.4' 3.6 ' 4.9'
- 6. 0
,3 TIE ( MIOC ) FARLEY #1 "W" SPECIMEN NUMBER.
- AT37 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-10. Load time records for Specimens AT45 and AT37 A-10
- r t i
-A f ranux et u-aim east Tsuns i i j l aI q. t A S j-l t =_ I . 4' 4..' TIIC ( MSEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT36 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1 l
- "W" CAPSULE resuv et u-arse anst vanns 4
l {9 W e 3.- n w v. n-I l .D 1.2 2.4 3.6 4.0 .0 TIPE ( MSEC > FARLEY #1 "W" SPECIMEN NUMBER
- AT32 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-11. Load time records for Specimens AT36 and AT32 A-11
rentry et u-ATes etsC TRANS i e a e I q_ t ^ t 3 g. =_ 3.6' 4.4' 6.0 E.4' s .1.2 TIE ( MIEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT43 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE ramuty et u-afst easc Tanns
.[ t e-5, ^ 51-e w u_ i s.6 ' 4.s' s.o 1.a' 2.4
- .s TIE C MSEC 3 FARLEY #1 "W" SPECIMEN NUMBER
- AT39 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-12. Load-time records for Specimens AT43 and AT39 f
A-12 i
.e E l ranLEY 91
- W" aT33 Sa8C TRAMS r
a!.- 7 n S *- n w F1-l i .o 1.a a.4 s.6 4.s
- 6. e TIE
( MIEC 3 FARLEY #1 "W" SPECIMEN NUMBER
- AT33 MATERIAL
- BASE TRANS CAPSULE
- FARLEY'#1-
- "W" CAPSULE i
ranLEY st *r aT41 ansE mnMs - a a 4 4 ~ ~ r 3-M w e-i y- .D 1.2 2.4 3.6 4.0 6.0 TIE ( MSEC ) 4 FARLEY #1 "W" SPECIMEN NUMBER
- AT41 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1'
- "W" CAPSULE Figure A-13. Load-time records for Specimens AT33 and AT41 1
I A-13
4 FAALEY 81 *W" AT31 BASE TRANS s a a g j a A 5 *- n w t-su s 1 e a e .D 1.2 2.4 3.6 . 4. 0
- 6. 0 TIE
( MSEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT31 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE rAnuy et u-aT34 saSE Tanns n
I e-
- a a
5 *a-w q. I i..' 3..' 4.S' TIE ( M5EC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT34 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-14. Load-time records for Specimens AT31 and AT34
) A-14
raALEYet*r Ateo Saar ThaMS 4 ~ k )~ z A 5 *- n w $a N-f s.e' e.4' s.s ' 4.s' . s.o .3 TIE ( MIEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT40 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE ramtry et r aras ener Tauns i
i " = _ I* E a n w y_ l .3 s.e' e.4' s.s ' 4.s s.o TIE C MBEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AT38 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-15. Load-time records for Specimens AT40 and AT38 A-15
r. n. / rentry et au-av4e east Tnans i !c g I w I e_ I f E t i ^ t S *- i a { w i i w_ ~ .3 a.: a.4 s.s 4.s . s.o TIE ( NIEC ) 'l FARLEY #1 "W" i e SPECIMEN NUMBER
- AT42 j
MATERIAL
- BASE.TRANS CAPSULE
- FARLEY #1
- "W" CAPSULE rastry et u-aras anse vanns i
t I-E i S *- a w =_ i 1 i I i. .4.s - s.e .s 1.a - a.4 3.s TIE ( MSEC
- FARLEY #1 "W"
} SPECIMEN NUMBER
- AT35 MATERIAL
- BASE TRANS CAPSULE
- FARLEY #1 i
- "W" CAPSULE
? Figure A-16. Load-time records for Specimens AT42 and AT35 f f A-16 i l y
4.: ramLIV et *ua fMDB - Mh! a a e e r. ~.. 1e s e S *- n w v_ .u _ ,L a .D 1.2 2.4 3.6 4.8 6.0 TIE ( M50C > FARLEY #1 "W" SPECIMEN NUMBER
- AH38 MATERIAL
- HAZ CAPSULE
- FARLEY #1
- "W" CAPSULE
'i i i a i t l i Computer Error - Curve is unrecoverable l Figure A-17. Load-time records for Specimens AH38 and AH40 t A-17 w
E ~ r i j ranuv en r mar mz a e a a g I 9-e a a S 1-w v. f l ".f - .9 1.2 2.4 3.6 4.8 4.0 TIE ( MEEC ) f FARLEY #1 "W" i SPECIMEN NUMBER
- AH37 l
MATERIAL
- HAZ CAPSULE
- FARLEY #1
- "W" CAPSULE ranUEY 01 "r MOS MZ i
g [ 9-I a S *- n w
- 9. -
.9 1.2 2.4 3.6 4.8
- 6. 0 TIE
( MEX: ) FARLEY #1 "W" SPECIMEN NUMBER
- AH36 MATERIAL
- HAZ CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-18. Load-time records for Specimens AH37 and AH36 A-18
.k 1 A.-' rasasy en u-moe nnz r 'e t I )~ ~ r a 5 't - n n- \\ 7 .o 1.a a.4 s.s 4.s s.s TIE ( MBEC ) FARLEY #1 "W" i SPECIMEN NUMBER
- AH32 MATERIAL
- HAZ CAPSULE-
- FARLEY #1 i
- "W" CAPSULE l
ram.tv et au-mess una .g j i i l l a 1._ e ^ $ g-w f
- y. -
1 .o 1.e e.4 s.s 4.s s.s t TIE ( MSCC ) FARLEY #1 "W" SPECIMEN NUMBER'
- AH44-MATERIAL
- HAZ j
CAPSULE
- FARLEY #1
- "W" CAPSULE l
i Figure A-19. Load time records for Specimens AH32 and AH44 A-19 1 .J
.5 F 6 31
- W*
aH33 - MAZ 4 I e,- a f S 1-a w u - s.a' a.4' s.s' 4.s' s.o TIE ( MBEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AH33 MATERIAL
- HAZ CAPSULE
- FARLEY #1
- "W" CAPSULE ranuty et awa apes unz 4
L I e- -s a 5 *- a 4 1 i "~: l f i 4 .s i.e e.e s.s 4.s s.o 1 TIE ( MBEC > i FARLEY #1 "W" SPECIMEN NUMBER
- AH43 MATERIAL
- HAZ CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-20. Load-time records for Specimens AH33 and AH43 A-20
c-. Fr: !.i .f I rentry et u-ame - ener taans e e u k )" ~ s ~ i 5-n w L t =_ 5 .e 1.a t.4 s.s 4.s
- s. o.
TIE ( NEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AH42 MATERIAL
- HAZ
[ CAPSULE
- FARLEY #1
- "N" CAPSULE t
rasuv et au-ams unt 4 i i e g. a a 5 *- i a i ~ $ ~. =_ i I .e 1.a r.4 s.s 4.s s.o TIE ( MBsc ) j FARLEY #1 "W" SPECIMEN NUMBER
- AH45 MATERIAL
- HAZ l'
CAPSULE
- FARLEY #1 i
i
- "W" CAPSULE Figure A-21. Load-time records for Specimens AH42 and AH45 l
A 21 I
5 g -.. t ramuy et -u-mot - mz
- j..
J L. f e.- I, r i' I ^ L 3.- n w h-3- u u- 'r .9 1.2 2.4 3.6 4.s 6.0 TIE ( HKC ) FARLEY #1 "W" e SPECIMEN NUMBER
- AH31 MATERIAL
- HAZ CAPSULE
- FARLEY #1
- "W" CAPSULE restry et u-mas mz g
t.- 5, E a 5 *- a w i v. =- .e 1.t' 2.4' s.s ' 4.e' s.o TIE ( Metc ) FARLEY #1 "W" SPECIMEN NUMBER
- AH35 MATERIAL
- HAZ CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-22. Load time records for Specimens AH31 and AH35 t
t A-22 i f
t i I rAALEY St
- W-AM34 M4Z I
"1 e e i I l ^ 3 *_ i M w .3 1.2 2.4 3.s 4.0
- s. O i
TIE < MBEC ) l FARLEY #1 "W" SPECIMEN NUMBER
- AH34 MATERIAL
- HAZ.
CAPSULE-
- FARLEY #1
- "W" CAPSULE resu:v et u-AMes Mat i
i l j =_ = ~ 5 *- a w a l i i a .3 a.e a.4 a.s 4.s s.s TIE ( MSEC > FARLEY #1 "W" SPECIMEN NUMBER
- AH39 MATERIAL
- HAZ' CAPSULE
- FARLEY #1
- "W" CAPSULE J
i i Figure A-23. Load-time records for Specimens AH34 and AH39 l ? A-23 i
? ry I raRuy en v . aH4 mz 4 {- s 8 4 a 3 g-n w s-l f w-f o s.e' a.4' s.6' 4.s' s.o TIE C NIEC ) { FARLEY #1 "W" SPECIMEN NUMBER
- AH42 MATERIAL
- HAZ CAPSULE
- FARLEY #1-r
- "W" CAPSULE FaRLEY si y ams MD 4
g- -a ^ 39-M w i l j 3 1 ~- ~ L .9 1.2 2.4 3.6 4.8
- 6. 0 TIE C M512 )
FARLEY #1 "W" i SPECIMEN NUMBER
- AW45 MATERIAL
- WELD CAPSULE
- FARLEY #1 l
- "W" CAPSULE 1
l Figure A-24. ' Load-time records for Specimens AH41 and AW45 1 A-24
4 ) raALEY O! W-
- sMt ELD 8
3 "g e-f a 3 g_ n v i ..e TIE ( MSEC 3 i FARLEY #1 "W" SPECIMEN NUMBER
- AW42 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE rantzY en u-sees iam 4
k7 7 51-n w y_ s. e.4 3.s ..e
- s. e TIE
( MBEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AW34 MATERIAL
- WELD CAPSULE
- FARLEY #1 i
- "W" CAPSULE Figure A-25. Load-time records for Specimens AW42 and AW34
'b A-25
id' s 4 Fin.EY c1 *M"- AMD6 LELD 4 E I 5 6 9 - I._ 4 A a 51-w T. =_ S a e a .s 1.2 2.4 3.6 4.s
- 6. O e
t TIE ( MBEC ) FARLEY #1 "W" e SPECIMEN NUMBER
- AW36 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE ranLEY 81 "M" 88M1 lELD g
1 I,e_ a 3 g_ w "i =_ 'Y l .s 1.a a.4 s.6 4.s
- 6. o TIE
( MBEC ) FARLEY #1 "W" SPECIMEN NUMBER . :AW41 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-26. Load-time records for Specimens AW36 and AW41 A-26
-l
- 4. -
ren zy et u-mas urto - 4 a 1._ 7 ^ 51-e w %~ o
- 1. a '
r.4' s.s ' 4.e' s.c TIE ( MIEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AW33 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE ram.rv et c mas una i
1 7
- 9-n w
y_ l = i .o n.e a.4 s.s 4.s s.o TIE < MSEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AW35 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-27. Load-time records for Specimens AW33 and AW35 A-27
.u
namn en *u aust san.o 4 g._ t a S j-i" n_ o n.e' e.4' s.s ' 4.s'
- s. o TIE
( MEC 3 F.MtLEY #1 "W" SPECIMEN NUS ER
- AW37 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE rumn en u nus sata j
4 I.,_ s a 51-n w q. I n 1.e' a.4' 3.s ' 4.s'
- s. o
~ _. TIE C MKC ) FARLEY #1 "W" GPECIMEN NUMBER
- AW31 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-28. Load-time records for Specimens AW37 and AW31 A-28
4' >y j._ FWILOf 01 *W" meat LERA i j g._ A 5 *- a E [ 9. l $d a
- n. -
l l f .o 1.2 a.4 s.6 4.s 6.e TIE ( fuMC ) FARLEY #1 "W" SPECIMEN NUMBER
- AW32 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE ranuv et *u" mas asi.o i
~ ~ r A 5 *- n w g_ o .o 1.2 2.4 3.6 4.8 6.0 TIE ( tesEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AW43 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE Figure A-29. Load-time records for Specimens AW32 and AW43 A-29
r F i I 'I r a rv.: u-mee no 1 a a a s g I m_ 4 i E
- a 5 *n-w r
q_ k r .9 1.2 E4 3.6 4.s 6.0 TIfE. ( MIEC > i FARLEY #* "W" SPECIMEN NUMBER
- AW40 MATERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPSULE ress.Ev e1 "W-em IELD a
a a a 6 T I 9-E a 5 *- e t w L 3,e' e.4' s.s ' 4.e'
- s. o TIfC
( MBEC > FARLEY #1 "W" SPECIMEN NUMBER
- AW39 l
MATERIAL
- WELD f
CAPSULE
- FARLEY #1 i
- "W" CAPSULE Figure A-30. Load-time records for Specimens AW40 and AW39 i
A-30 i m
r FAALIY 91 *W* 200 lELD -[ J __, 9 - I* 4 r n i e S 9-a v i v r .N-i I r o .9 1.2 2.4 3.6 4.8 6.0 TIE C MSEC > FARLEY #3 "W" t SPECIMEN NUMBER
- AW38 MATERIAL
- WELD CAPSULE
- FARLEY #1 i
- "W" CAPSULE rantcy et *u-ause mm 4
t "e i l,* 4 6 m-7 ^ l 59-a N i q- .9 1.2 2.4 3.6 4.8 6.0 TIE ( MSEC ) FARLEY #1 "W" SPECIMEN NUMBER
- AW44 MnTERIAL
- WELD CAPSULE
- FARLEY #1
- "W" CAPRTTLE l
Figure A-31. Load-time records for Specimens AW38 and AW44 i A-31 !}}