ML20092J103
| ML20092J103 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/31/1995 |
| From: | Auerwald J, Rishel R, Williams J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20092J085 | List: |
| References | |
| WCAP-14315, NUDOCS 9509210330 | |
| Download: ML20092J103 (130) | |
Text
- .* ::T.'..a.
. .**. . . . :. ' ' . . .;l'?; *;,;. *: ," .' ,*.?..- ' .
.y,- . . , s
- ,. . ' . - , - * . ..* ' +. ; V : ; o..'-'.' ': .l-/e r '; . l- .:.. V :. .. -
.'a:'.'*Q*'.',,...i,..*".'?.;
~
'?
<
- l%, :*
- r. f . ul ...';4- ~. ;; ; v . :...; '.;~., T y r; ; . ':o .-. . '*. .;b, "g,t* l ' ;^,l ^.': Q^.
i ,
- , :D :, ' < ;; y'Y,,~- ,i;. g ,'.s '. , ,n* .) '.' r-r v y:. e... lK - .. *-* '; .-:.. .- 'J;f*y ^ - -
't.l . , .
. : ^ r.. - . :. % y (. : . ;. .: {, . .;;...).~
. ,a. . .. ,"(, ;;
s
' f * -
- . f. :
, : .:. . m.:
.. , f,
.>..v,;,,
. : .d
..:.y a,;; (*:.
- ." -. .: p.g . . . : ;. .. . e .,
i ,,',,.;. e ,
.n.- ? ., .g ::,,' :
. .: . . y ,;, .f.. . ., . ._ .. . ..uro.1 ;.,. ,y.. .ws.,,..,
- (, c% t' 4.. . . U,.n. u?'^. l g.. ...,..~,.'."
- ),
v 4.f . .... it.t.:
~a, ..
, . s...
...,.y..m.
lK y
}: :.
.p
. y
.l><;. , .:.:
. . - '. % ;,5.. .
g n, .
' ..c.'. ,.s
- n.4 p . .-
...,..r....
>_; .4 y y '..:
- ka.c5h. .f. .*k '.,Y; f'rll:_N,_ ?
- h. ;;h . 1s.
- 1. . 1. "l,Y ,.If. ':e..Y, s. n$.
s.. . :A ;',.k.l,:?. q'.*f. ,l.'f. , . f. . .s*, '. s. n...,:,. .f.&i..,.,. Qt , i if?, . :.tu
. ir '? N'.,';
4,,'
. . .. .e r-e;.:,h.:Ul ,.?. . ;...'?. ?
e s .- .
s .
p.;
. . .h, -
r, .:,.....:. ; . .?. . . * . . ...
',. .y L.', ..,. ~.c.-u: ..,
3 m.. . ' .i ' -p. . ;
. ~ ' . *:. .
.- . .... : :.f.. . r .l :L. u.
. " '. a . - :-"'.....s- !
-r
,.:p :. ' *: ' : ' ..'c e..;,<
..? - . ...! *.;O. W.;,'.
9 . ~: ., ..
.7, . ':.. y. . .
' ' . -f %.t. Q 1.25 ' lp.v.
- . . ';n
- :,,,,3. y
,- J , t n _ .' .i.,
y .,-,. ,c
-;l e.'_ 2 . y, ;q uA t*. .,,.y :,.. -3 a . . < , . ,e..
' u .~ .% ' y:
- 's 1 ,;.,
. y W.; ; ,1,1 . G :':: ., *:;, .& j*f, . ;.: ;'.h',".k.
. sA;4._, .. ;
e
. Q. a.
., .:,;.. . ;,, co.*' ..s. . ; , v: 7..s > .; ,' :.:; -
.,t.,,;.,;...3,. n;.,. -
. .: e.. .
..Yl.'.,,.r',.'^.*'. e.'.
<A.....,1*. . , .t s *: y:. ;! :. s. l.'!fh . .*"Alg, T-
' .N )l*%E.. ..?- ..\ i ..
.. J.f, .l - l l ; V - : ' . . : $.".'.k. . '. T
- .?l. ::t.\ % !.
/, :
.: .' -h'q' i,' .-D,' ,5? '; ',.?.- .i !*^. * :* .'* :.i:; Ly
'.'Q'Q.t *:?.* ;'
. M..N. 'v.;> ?.'-2*;;'.- :. Qf.;;W'? .},'
' :*.' . ._.;jf;,;. k* . .).'.;.. '
) l,' %.G:f '. ufl "l,3;
- k?l,t. " <, lWl:. ',;,.y.Y; '. L A.' ;..
. , . .. t_.
j.
N p.Y f.
lQ', jo. .%.,,,.. b, l * . s
, a 's!.'. , ',t
. ' ', % l
'. 'l>';'*'.
.... .,; A o ' ;.* y d.:%,..***:'-
. . ,. - i
~;.'."*.'y.'.. . . i f: g;f.:%'~;{p . , .; .. .
.l .' ' $ ;.,.-'t ..--:g: l<..l .. c.. , f, . k: . e- f.h ' Y.y :.~*
^
f &.~ ?. * ..b
=',-
b ,:' ?. ',.
..4.r . ,:. t
- .t* - n;:4. . . . ,i *. . . .s..a ; ,'
.- -..; ....s. ; - l q.pv'-Q
, u. '.. or.:* , . .:
. n. %,T
':* * ."'D@:,
^ '"& . . s .+ ;:. :1-
.s,.,,
a ' ~.e;
,.:'.,l -l,l .'s s,:-
,.:. .s
.. q: n.,.h .-^T e gy . .o n y y ; M .2
. iG. .. ', . ;,g,A. . n, . r * . :
- J;. 1. -1.j .l.W.". .-
l . " ;.,t .\.:,.- s < .q, , / ,;> -
., '. l ?,f, M g.r; ,.._,-P ?. . a ., ,_ ..
. . ;;..' - s ';. <> [ .:s .i , *' {;..; y. , ::O. .'.-.* g./.&, .- .-;':,b. - :
> r .'
, c.; :....';.
y k" .. . m.:% .. A :- s? - .
. ., a q. 4 ::y,
>L.,, *i . .. . ...s.,, ;; .,: g. . <:....a r ..s.
, ,.. ;; ,..s..
. ~ ....,y". ...
/ *. : ' . : ,.. .t '.;;s..%, ; ..,,#: .>. ...*e,.,..'..,..
, .s:: . :.' y' pk . , . ,
..:.-s..;
- < .l:.., . . ' ~ -* *; ; '-
- . ......~......m.e
,.v..~..,,..
...> as * ~-;. ~r, 1
. . ..< . , . u
.:, 3; . , e ...e
- a .s. .- . ,~ : : s .. . e .<..'.,"..,4,,;;..e *:p . .,.
- .,.<~,,',:.......,....n.
- s. ,n..,p a , . '. ;y ;. .
.u, -r....
- : i , r ' A':' ',..y.:
< : f ~
,.9* :.,. p'. - . t, . . . <
..a n.** c,.- '-
- < t..
-q . ... i. ..s
..n'. .:y>' -.%. ,,.. . . s .,' V: .. ...c....,.;
, . :.s. .-:..m.V' :;2 .:: *;!. .. t., y .. .- s
.c :,n . . ., ; . 9.y.~i . ~.
e
." : y..., , :, * ... . . , ,:.:. .;;. j.- ..s *'. -!,'.,. ...' ,.:'
'. -. <.,O: ? v.
,..". .- .g,, ;, '. v. y V. G; .';p .: .;; g a;L :
- , ;;b f. ' ;
- :* i ' . t 4. , ~:*;y : y,.'n~ . p' . ;
.. ..:. .,; t .;;; "...:i. . + ..':" V ..' :. : : . y>..s,-
,:.. , . . _. , ., . ..a, r.,..,_e
- ;
- e
.>.)
..:\. -
o.- ,1..,,.,;,,:*<s;._. o. . s- .. .: t. , o . . . , . ,. ...<.9,.,,,; ) . s.x, . .
~e.u...., ... ..,;>..e.. .. .. .',:* .,:...,,.,
1 a : * : . s. <. ,.': ....n...._.
W i.. . , . , ,
o o ; . -
.. .e a- ..: .. i. v,
= - .; . s .~.a i . . z'...:.,%. ;'? . *<:. .* *%,%.*:..,...._. :~:4-.* .-c.
.e .
F. v. .
. . ....;.",. . ' sR . 'y.?.vt 'u y:; :::. .... y..;.v.&-:. :l,1.'. . ; . . ;).at.e.c.:,L ~>n . . :
- . s .?.'If-..V. ;-f. :-
- . ' ' ' . ;;. l. :
- v
..p:
- ., p. ., , ': 2 .*ir::#'.-
.- ; a t.
- . ;' [:~ ;> G p ':"4?? ' ta& ,2 .s *: - "*
llo, A. .h,,y,p l.; * . : ; . ; dev . ..'.-f.f,. . ', 'j.*f. ,; :** \
'?
f e = ' ~ *':~.:.: . .~ ii ' i' . :. : +. . .
9: y:. .i
. '. .ll;..[l. ' '.l if,'N;h';f.::;.:, ' {.3.',, _ { ll * ' l $f:&,.. ? .::, .n i: - :a*.
R k
- ll.,f::h 5,..;f......;... . ';: g;.,?: .l,{;f '.?.'? l ._* . 'l.
..r..,.....
, . . }f . .}l ' .
s .c ,w", .*;l&s. f):
[d '.h l.".'h .?y$,Yl'f. ,
.). I.7,h'. h;\ k.?'Y*I.Y$,?'$$? ' 'Y S:Y Y*k.l;., -
- f . ; .: ~f,
.p l
.- ?.
J. .y ".< '.:.a . f-' .S , ]e a'.Y,'p. z';. , . /... i.. -,F;-:-':.Y.e. ' . . . , .. .
- s. ;y t. _7 L*.5,6. v,k's . . q.-(> ' .: . . ~; > , . b.:,,.n '; . .[;y.'.. .'.
.f)f , ax t r. ; .,, e., . .V
- .y .
<,....+.~A,.3*
, ,..- 4 4
.s. . , i., :.. 7 , - p . : . . ,%. . 3.
.r o . . g.
4 W".., ,-
..,3.',.r..,s*
.e.. g,...g.. ,..
b 5 .. s . ,. -' ..
...n* , . . . ,'.s<..'..vc ..nr ,
- d. . . ,6
,; ; . .; ps *.,...u" !i. - . / : '.4 :.l , r ? ,.,,:. C.; *; :1 #
'Ya . ,.e f ,e * ... . ,'? it
'i . i. , ".,a t ' : je .q '
^",;'* .-; .: .; .
- . %p!. 'n.s -l.'.., % W,s. w .
- ?.;:* ~
^
,P,
. . '.r '.:. 4 .. :.: .
s.-
' .k:
.' . . , :.;, ' . .x .se.u, s.... ..'e '.: 1.'.*:: , .:-. ,'; .% :. c. 'i_.
. o- . o .:
- :. . . . . f. . e . %.. ...q .
\
W. .,u.>.
. . ...;c ;;z.
e n. ~9. .r. s. . . .',4, - '..
...-lu,...,;*.,_g.... :9,c.. ,.* ..* , , ., . , -. .
- x. : , . . . . .- , ..s .-. , .*.v. :,.:; e,,4 hr:::s - m
- .I;,,
., y .. :. . ' . .. ;~ - :. .. A.> r,;;
.',il l ) <. .* 0 j.l ' . l,gl, .:,..-, {: :?; l.7 e O,,'lj,. ' f. , I&' th a, ,
7., s n.
..kl
. t
- l. Y- ' ^: 'l ' : . : f , ,.
i.; $ f;*r.*:
. i }!' ..}, ;; , , , ..; ;-l.j
,*.Y.
{, .h,'ly '~'. , ^f,,.'s ll^: , .-
.','.9e.... ;{.\ .;.f.1;.- y[ 4 j!! ; -Y;i
.,: .c,.,,..p,',
.. .> -^ ;
.. :..'. , ;;:.' .. , [ .h: .. .: ,i...'a
, ,o;;..* . : e'., -. "..;y
. ,fe .;. z*,.-.:.. ,
.g g.. -
-y : '., ,.;. .;; :v,I.. 4.< e v. 4;* .'. > ,r ,l x .,.,%5,,,....:v. .,: ;.-.,s.: . .
. , .. ;; ;.. y '*..' { v. .%,-
.. - y:,:. g#.s.
.=7'.,.*,. .;. ., . . ' .' :.i..,.; ; ..s.,.~...- '. ; y : , .,,.: p, . s .y, ;4 *. i . 'f.
,. i
. . : .; ,e',.
, ; . ' . 'l.% ,, i, .. 5 g: s4 !y .- ' ,ew \. ' ., 4 ; . .;~. -
. l, ,R .' , * :: * ' ' ' $., ; ,', ., y{l./,,
' ?'].;g'f ,. ,,.D.e, ,,**,.o, .
' ' ;p., ., 4 ~ ..l 7 1- .
- l y , , . '
,('
g/.. f . ,- ' .' [ . - 7I l .l,.].'; .' d :. . * . . . '
". * ' I;. . /'.l. ', Q,',o
[.,,,*},..,'.,
.[.}
. .'.y,' p ',.* ' j. g ,,'. l j' ',.,. Ag
- 27.-] ... ;, ;
- f. . . l ," .s s.t. ,., P:
..4 4 ;,,r; . .i g'.6.' .. ,.,i.,;. e< . , .,i ;;,,a;..S g'i...,',.1..f.,..,. i:
<j i. . . ';'., Ig 3' /. .' J., ..M - . .;< ,s '.
.; e ." ,t.
...y_..r. * . . ,' 9 : e 4v.~..:'..**.c" .
f 'l/,js'f. ..)
>. ; j.. ^ * - , fl *Q [: . ., . . . ',,-
- v. i., U, .
V,." ' : : :; ' ,.. f."4:j,. : ,
i::WL.. .'.:a..' A.: , ' 's i:. . >. '.y 9
' ;;.' C W. l'.QlL- { ?.3 . . ::; ,, W@d.' - .:W: . : . .n.. -?: :df ..v:. :s .i L . C. 'lC. % t . :t . Q.%.m 4;
.s . . . . . . .s .. t. .r.
. + ,; -
- ., .G" s.*- M.. ;. .p;.f. !.. 1
~. :_ . , *
- llf*f'f${..'$h',
.. ; A . Yh.hb., lkfff. .h, b f*
.I : .
- ;?,: ' ?' l- '.'
.N!.R ry f . ,9 y;;:p.:.:U j.; , . ..
V';klwy ? .'..'. p. Yf'.ng' .) Y.k *. *
- $. . :. , Y;
--.. G . .- . . f. ;* *
%. ,f' s. .h,"A'
- . ;4
- :. . . ,.:. E:,, :. 0. ' **,.*[P.e * ., * ;., , s ( ;;,. ,.t&' . *W't;^g'f ..% ;n
- sjf(a;' 'A .l,
- ' X
- lr . ,h',.; ,l :.p*.
sc.
Q.s:<,pls.. ' v .U: 3 ,
- lp'e 4%l..Vi .U % n i
. . f. h'.I', ' ' ;*y,d, nh.TP.
. i.
{l1.g ;;-
' ,t .',* ^ ,,l:; ... l4: '4.,{.. .' ...
. > . ,s h,,j' f. 5. :Y.' ).$'$ h; ~;kh
$. .;.D. :
y
,q. . *V -
O~..!'-
,,"}.?Utb...i,:,.{.&.s..W:,.&*
t
>p *.- -
/T' N.;.ll:f. s a . . As a. , .'#
. d i; dt ..>
, . .i..e.
. < * . .,p&w n .v.. .,y,*'?"i v; r, ,; m...,v,....s.,. re 9 .NNl:{i -1 . .s,-
- .:; r-
' ...,&,c.y&' ).i '/,t':e ". -*<.. :;f.:, '.she;,..%
$;'$v.'W.t;..;.:w,.?=
t..*'
,; , , . ., : . . i. ? /.,.by 1"* , se :
y .Q qlh * - ..h,!f.:
Q; tl.&..
.c =t .,.,.: f.. . ". (8, . ;L'
.g.
09.0%.e .lo. I;': * :I
. .. . . .r... ;e s. ,"f-q' f.- /- '/ ya*:'a'd.24 '
,..o.
^
- .. : .; .- ., ....a.*., . <. . , ,. . .
.s * :a. - . h
'- ]' 'l - #)
'..$' .[?hh .,...-.f. . w . ./.f, ;* * ;I' i.
. :.: '. s...: , :
.'.b". # '#
. .: - ...# : qp ... . l' n .u '., .. ).'l;i,-['.. ;lae \ * '.f.~ .* l.:.,.. ;. < 7 . '.
.. .<"m.- <. . .
.g',' :,,.
. ? *:,.,,,i'W,: ".. %.k.n . l..#h.f
....,s;.*-
...v~ e.
' L., .$ ;5 f; ,f ,M ' ' '[;.'h )";.
. .f. ' ; ?,?,
- } Y ,.; f.t: . :,'. . .A. ,, s;C,..
a .~; ;;'.Q * *::d%' .ci:'. ?sc : . .:f. .
- 7 .,
- ,M .
.)^:;
.f. .d. .: .. .],,y? '.-l :y.Q};:.z ** L.j;.. ..Jy,p;. 3
. .,0?,
.. j .;, '; ..n'b
^:.. . . :,,.:t,5..; r .', .. . y o h.s:
Q. . '.. .. ., v.'.' -'; , * ;.~. i. .'Lr.~. ':.;f* U ***'; . ;; .:i.i' :'4;_ }*qyl' % !a:',*:
- l
- . . :' - * : : ' : ;*
- . .l' u.
,e ..$..,.
.. . " . .,c ,.. ~.-:.., v....~.,*,.~g.:e a..:s .g. . . ,ei',y'*
r
. .,. t t, ,'g' . ' ';. .qo" . .: ,^:* . '. . :l:.'.?.r'< ; ., . . -e . ;. * ;. ,: .. ., ,*' . . . , ., . . . :. I
,,, ',.:, 31. ,, ;' ,c. , ,,:.. .,...%4..,..s.Q,.9 ..;.,', ;- - . , . .. . . , , ~ . ;,.' , ;: .i. ,;s.'m.. .,..., . ;. . . ,. ,":
. :,,,; . .y a. t;pa,;,.; . . %'."...,,.v,
,s .. .q..
r
. .. ;.~.n.i a. . . , . ,; ; ;L.
q a. : 5, :. .;- . . v.. " :t
.,, v,v.
- ' .,3.
'.# %. n. . .y: ; . '
,Y l / I I .D.) s.h,',s '.'i.
. [ . :[. .' 5, [.. j;* .' f. [,]^')t,. D. ;., : ./.I.,%,.,,...h.':
h ' .,. . .'. ' . :. .f( *I ., $:f ',I, ,' c[ . [ j N. ; ,h,f L - ' ['
b, y$. ,' c
,~ .,*: j .m; x.
.p[ :, . . ., .
y c .=a:-b ;. .,.:. ,...?.. ; ,,.6,
~..,w..,W'
.u:
. ,13.
-..*t'..., . . . -**- ..s*e: ;d .: t . -
' . ;s.- **<..t. . .,* 1<: , .. . c, ., ,.;. ,. , . , .':. .- ..,.:. :;.* :. : ..c.e ": < . . y , . .(:. y4 ,5. .:;.
1
. : @dv... .....vm.." 9 '
e .n. .s*.'<,t.
J'~.* ..ti *+. :.
. , . ' . , . . ;' . ;. *. .g3: ;,:..-o.'. *::.....'.,1 .e1v*.
- ;. s.'
- .s..t L. '.. .,.y p
.c,'....*::..,.". ../.9..,...,p4...3. . ,,. : " .s. .r ,,p y 1.9 , . ..h J ,& . y+.1......,:<:.....,.t,,.,
c
- 3<:
- . ' ' 5.g q' * ,. ' . . Q . t <.
. 6*
a r.
3'
. . - ,.q,.,.. t, - .. . - ,
- .;*4 .;f,,;t;, 3, ,*ts.4w.- (c.t.*.;
.e ..-,
m' c ,,,p+. ' - ,. pu:<,;
.f' ..:.5A 'n . . . e.; * .. d5ef p;. ;..iy.h r.,, :
- . '2, ~ y.Y' .L - L 4. '..*r ? ^3;Y .y y '. n<es .-
. :li 4 2.g
..e..'..
.- , . ; , ; ,. , .'. lt .
.? ,:
-,>..v.. '. , }' '.h., .,*;',<a " j
+
l . 24 ','. ., \ . ' . V'f.,.,:s' *
- .m n
' ( - Y, ; *% ";- ],t' ".f. ,.i. -'
m: . .'r *
'.? ':: * , ~ */ .: 'T..',
.c . * : ., ,,f , ,.: * ;;' . :)*l:, ',':l( E.: ( [ - 'ik. 'll'* .4?. .%. r ;::. .
n, I, *:l, ; % . 4. *. ? ;: *..
^
. ; ^
e .
. ..i- .. .z . . ,. s n .
b.:. ,l^; .
- ,' -;. y . _
- , _. . . ,.,. . . g .),.'t.M,,
< . .'lg ?}: 1, ,.
%1
) .
. .!. ,0.p
- ) .
> ,[ :,i{,
o,. ,-
o,s' ,.,.i
. . ' , ' ,. . *;; , n. dt; 't 4:[' %
. ,. ; .- ..- . . .s s . c; s . , ,,, [..,.: . :.
,,f..... . * ..u.. . h f*.
,. , ',.c . . - . - .
- ...k *.Y b ' '
..... 4-as. . %' . ..c. . .. a - ,- l, \ - g. p y. E*. , - .:.
. l hp,, ..j:(.
- .2.. .. c. . t >
- .? . ?!?.f O' ...f '{ h. , ;;,. \ .. .':, ;h&'.,' * ; . 'h. 5.h .fll:W.l.h':.h.'t'h Yb'..lth ,.,~ ' ?:' . * , * .f 0'kh I; ;j Yf k ' ', *
]M,,'$. T'1'h:q Vnb@<**)
_1 i .9/W.s.'d).,Y.. . .. .Y6l$; :,[0.. NM,1.fy.7:D.
$,y'/i'N. Nm ', r*3..s.p' ,v :.$.- @4 M' M,.jp,.
s.M. ., lg;r: .';'k.*.&i?2. .[?.%,YM.ip,D, . . -;, *a ') :y* : '*' .x _: .S.N' <-? y':?.'@r6.[P l
'..a;., f
%. r :- h:!n.+ 0. $ \; ".1 c
'1 - ts
<&b*
- 7.r a*' sc e
. ;; .p.n . ' t ;:.'$*% e%.:. ' '. > ,Q.,yd:*M.h' p: ~ nnO p+ t .0 w?
n 4 ' t'D A.:h sus.
s.; . _
,, q \ ;.w.g. - . *e=.... +. ... : Ae, . .y '. ,J v'2;)%' 4. h. .Q . .L -* ;,c 8, :.~ A;.vQ' . <*.. A ; . . 's . -,.~.;.'.e.r./.". .
' .;; i n
- r r
- . . 1% .'-
- c.l' , . r. >
..'.', E'. , . ., G.4 . :,.',. .;'t,$'W' w ty :. ..
- ?.k5
.' h. b; .
- . -l
- E
'): ?,.,,: k ~*'
- lQ)'. :
-h , ,
. ., , . y;f. ,-l yy. . .. es' t.;W!L. ,;~}. %., * ' p, a*;yllh:% v & l' Y .. , * :r qy. :,f..*._ '?.3,;F, :;.-. l- ',
?t
. m c . .\, . . .: .n . . t. . , . ,e :~. e ' . < . t" ),yb e
f,l.3.,M. ....y "Q +p '.. -
,. .'y:.>.ty,*y$ y t ,% f,,%.} ,, *
- g. 'i- q:p.: ....} ',' s$ '*&. - o ,xy' 8. 'b:y;.:p?, < > .b,%.Q,
- s. a ' ; &.l: syg Q;' ;Q -l:,L
^
3
, . . , ; ,T '
't' : '::h.y,*'j f{*y &y;l' &I.{l_ i \!.'&,l; /
' .( s ' :.**'!.-E '.f .,%.
s & *% !.l.'l. '.. . : ; '.~ l; &
- Y
'.l,fI.?f !l,f., s'; l.T.}lff**) 3';; j.jh'k.,;g' h l. "*;
- .. .. l {' .{. f ,.; ?
^ ' '
l '
- '
- p . . . _ 3. . .
' \.,Qf.s .Q.l. o f.': <c,,*,Q, .f y,: t g' ?,.t/t c.mg,
,,t
,_r i.If n ;. 4 ;,.
, 7 p,,g4p'.,.;.
- . o. ,l lf. 3Y),
,:;;..R .l l p,. 1f. ,%)[
- i. .f .N '..?t_ 'r. ,q ,
- f ; ' $d . g ,q'[
- f , ,.dh L $ ) -.
- $l ' . ' , . ? ' ih f '. k' !
i/;.Gie ,r. ':. efM . . .". y^e,,
Cy . _'. 7ln;[f;4yy , A.t.-
t,...
. . %i .
1, . $;, ..'.,;%yycn p l:;..p,u,j;,@ifi,*;l.i; :. "*. . -
l f l .* ' $ *
- k': , y f.' $Y;l?:' ')~s
.. . ; 4 . ;. .
,?? f [.$ h @W.13; ,f:';;.!?f{ f js:,.;
, . . f' -e l;' 'h'h , . @. $. f.
?
. ?; .l h h. 'hh. #f. ,
' ~~
....E*
,', ,j '.,J,;
. . , . . .. i [. .
', ., 4 -
- ,', -. . [..', '
. - l
,. 'Y;f)..: - '.+: ,.,l+ . ,* .* ^ k . I ' ,, * .
. ' . ' . . i ,.: ,.I g *5k.'
- ,.
- .p* g .,.,i c.
r ., t.' . . 9 < .,; Ki....Y.
$\l.. ' ,y'Y* 1 h' 2- <- < .
s * . . . . , . .4. ,, .. .c w .s n;g .;,,, ., . ,m,.
- >p3, g'i;f, b. q . . 2.,(u
- f.c :e,J-
..,. e . .t $
.:s. ; , ,
- p4 ..,q,:..e. . .: : +. <. ; t , .7 .,: . m, t i.g. 3,. ,pp.y~
. . , , ,. ,.. fAJ . *...>..k,
. . . .u . . : - :. a
. , c
...u..,
rc hg: .;' ,- . 6y,f.,'.+... , . .! : n ,, , f. F.. .M, d. /4. u4 .,.
.. s
.."l- :4,g*.m;: w ti. ... .v>".3,.,y
. . .. ;,- ' a,. '; .p: .. < .:.m. ., r. .g. cy, .;:%. :.,,'s/.y.,:,n,, gL. . s "s :
- 5. g .
cr. . y f,*g 3 .,.-: ..g, . , . . y. ,,,,, , .
,g,,, 'i e ' .p.
7.;. ..;
' ~
? '
. , ;;:,y' I ," . . y '; ,
'^
., , ': l. " ' , . . ' ;hl' *
.pf: >
. l.[,c f. u-. h I* k',:',*.' l,
':.N . N.fk ? h'*;*'Nl-; Wll'
^
. : . , .; .. . ?. .-- l ...g,. ' ,. 9,g.. . . .. n.y.n f,.-'.g'.,s. . ". '+.^ ' , , ' ' . ?.), f!;%.,f , ' ? ,' . :' . . .f..,,:? l"!
.. ....a. . . . , . ;. .,.,..c.
.J..-...,q,.......;*f
. .. . 7'. ;.; ; .V,"
S.**t i9 . !.. s' ? ,r,. '.<., .; , . ;.5.c E'
. L .'?. ' .',N' ,' & ; 'l:' I; f h .i. .'.f),E
., r.;
.,pi
- v. e. eb .. : ,. .t- ,.* ,...,,; p
. 3
- m. ..g .,,4 . . g.. , . .;... j j, . ,, . ....j.,, ,(- sj.,.y'..-..?..,.,.h..?.?.;.,g . ,., ;,
gQ'.* .ta.- . ...i'..,,,.',,g.. .,,
',.,..f.y,r,,...,,... y
.e r
,g
, . , ; ,. ., f .,. .
,4 u.,,.
,,*..: ., ;c ;+
.....,,.eg,.
,,.,.*..,;,.',;y.....'. .
3,.,9,'..,...,",+:..r.
. .,,o,..
,' p 9, e.
j
- , ,. .... - ., .,- ;',,. ;i:, :. .. ,* . .v.,,:,. *.4',*,.. , '- ', ./'.,- " . ., , ~
v.-.- . . . . . , . . . . ' '..N,:.,.,..,A...'q', . *" . , ;. . ; ,*t
....t,",.9,.: .e..,f+...,,
.,.: ( , ; *.
y%. ; ,.....,f' . .i .-s$ . .*.- .<...,..j..':......:...,-
. . ,- - g 4.. . ,,
- . s;. * ,: * . .,. ,r , ;.;.f'
- s. 4-a, .
'..,...,t.,,'....'.. . .n . ::: . . .;, , a.....,,i...,.
,n','
- v.$y% . og ...v.
.' t . . ':. -.. ,.;;- ,: . : ,m,..,.* e.-.:: .,'=. *,
.. * - '-. . '-'i. 4,,,...',..>- ,'t .1.;.
. .. . :t*, .W ', , .
'y. . - ~ ,,;la!,.;'..n.**.....% ,1,;,* :.; . .y ,.- c..-[3*.*',!...?.
+
.c.. .'..w*:
6 .; - - .,
- - -. ' .
- .: /.
,,;.s,, l.,'; (. . t.f m..'-
" , , , . . *: '.t , . ,_...:... , :;: , <...,;,.
,_, '. ; . ? .-2 .',...s....> .y' s . .:. .m.,;
, ; .. .r, +.
. . 4 y. ,. ..,'.;q ._ ; ,,y ,'.,, . .y s _., 'r; s. .' ;;. ,;\;. : .;,:*
, . ' <.. s,. .; -,-*".;.., ; ,'. s, ; s:. :,; , .M:.& < . ,;.,;., ,. .t .pg. ,. . ... .. ... . :.%'v; ,,3.
,, u ,
- . . 9.
L;.,..-
.,..*_v.,'.,..f,,.t. y; ;
- ,... . . . g.
,s: ,- , , , .o.... .
, ,'. .* ,. :..... ..r: ,. .:,.'.e,.,...v
. .. . . 1
, . - , #.s,
. - . 2.. -
y s.
g s - . . ,3
,.p..,. ,,'f.-
9509210330 950919 < J. .?. ... .,,M ' 7...s,. 5~ E % t;i e ' 2fL'bC l N W ?n S. A. . . a.. , .\ .1 d... ?.
7_ FDR ADOCK 05000446
- ' 7. ..- . . e ,o ... . , .
.t.
i . i ~4'.
. ...V o . , . .
P PDR Ve ; {l ":.'.n ;:h .6 du.i:{b i;" u>@g'.';; ~ ' r. l .lc :Y'aT: , .?. . .' ..;p? . .. '-L:; . .
e.Y;%W..!:;a ":.G + -l:( t.'a y.
,..~,
Westinghouse Non-Proprietary Class 3 i
l l1 I
r 0
?
i' T
.l} }
- Ai! 1 t' ,
.q WESTINGHOUSE NON-PROPRIETARY CLASS 3 l WCAP-14315 q r
t i
ANALYSIS OF CAPSULE U FROM THE TEXAS UTILITIES ELECTRIC COMPANY !
COMANCHE PEAK STEAM ELECTRIC STATION. !
UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i
R.' Auerswald }
J. M. Lynde ;
J. F. Williams j i
July 1995 l i
r ,;
Work Performed Under Shop Order WFTP-106 (
)
[ Prepared by Westinghouse Electric Corporation ;
for the Texas Utilities Electric Company ;
Approved by: /
R. D. Risht1, Manager ) l' Metallurgical & NDE Analysis i
i !
I WESTINGHOUSE ELECTRIC CORPORATION i
' Nuclear Technology Division j P.O. Box 355 -l Pittsburgh, Pennsylvania 15230-0355 .j i
O'- 1995 Westinghouse Electric Corporation !
All Rights Reserved - !
i
- i. }
.; i
'= t
~
PREFACE < !!
r l
.I t
'l This report has been technically reviewed and verified. ;
e
'I '
l i
Reviewer:
-l
'l
!. Sections I through 5,7,8 and Apper. dix A P. A. Peter 0A 6EN
)
Section 6:- S. L.~ Anderson ' N.[ OJH'o D oh 1
t k
I i
i i
t k
e r
i h
t I
k i
f i
TABLE OF CONTENTS Section Tillt East . ,
i LIST OF TABLES . iii
' LIST OF ILLUSTRATIONS ' vi !
1.0
SUMMARY
OF RESULTS 1-1 l l
2.0 . INTRODUCTION - 2-1 i
3.0 -- BACKGROUND 3-1 f
4.0 DESCRIIrrION OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE U 5-1 5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-3 j 5.3 Tensile Test Results 5-4 -l 1
6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 i
6.1 Introdnction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-6 6.4 Projections of Pressure Vessel Exposure 6-11 7.0 RECOMMENDED SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 ,
8.0 REFERENCES
8-1 ;
APPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS A-0 l i
i 1
l l
1 I
ii l 1
LIST OF TABLES Table Title Eage 4-1 Chemical Composition (wt%) of the CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate 4-3 4-2 Chemical Composition (wt%) of the CPSES Unit No. 2 Reactor Vessel Lower Shell Plate 4-4 4-3 Chemical Composition (wt%) of the CPSES Unit No. 2 Reactor Vessel Weld Metal 4-5 4-4 IIcat Treatment of the CPSES Unit No. 2 Reactor Vessel Beltline Region Surveillance Material 4-6 5-1 Charpy V-Notch Impact Data for the CPSES Unit No. 2 Intermediate Shell Plate 2
R3807-2 Irradiated to a Fluence of 3.28 x 10'8 n/cm (E > 1.0 MeV)(Longitudinal Orientation) 5-6 5-2 Charpy V-Notch Impact Data for the CPSES Unit No. 2 Intermediate Shell Plate 2
R3807-2 Irradiated to a Fluence of 3.28 x 10'8 n/cm (E > 1.0 MeV) (Transverse Orientation) 5-7 5-3 Charpy V-Notch Impact Data for the CPSES Unit No. 2 Surveillance Weld Metal 2
Irradiated to a Fluence of 3.28 x 10 n/cm (E > 1.0 MeV) 5-8 5-4 Charpy V-Notch Impact Data for the CPSES Unit No. 2 lieat-Affected-Zone (llAZ) 2 Metal Irradiated to a Fluence of 3.28 x 10 n/cm (E > 1.0 MeV) 5-9 5-5 Instrumented Charpy Impact Test Results for the CPSES Unit No. 2 Intermediate 2
Shell Plate R3807-2 Irradiated to a Fluence of 3.28 x 10'8 n/cm (E > 1.0 MeV)
(Longitudinal Orientation) 5-10 5-6 Instrumented Charpy Impact Test Results for the CPSES Unit No. 2 Intermediate 2
Shell Plate R3807-2 Irradiated to a Fluence of 3.28 x 10 n/cm (E > 1.0 MeV)
(Transverse Orientation) 5-11 5-7 Instrumented Charpy Impact Test Results for the CPSES Unit No. 2 Surveillance 58 2 Weld Metal Irradiated to a Fluence of 3.28 x 10 n/cm (E > 1.0 MeV) 5-12 5-8 Instrumented Charpy Impact Test Results for the CPSES Unit No. 2 Heat-nffected-2 Zone (llAZ) Metal Irradiated to a Fluence of 3.28 x 10" n/cm (E > 1.0 MoV) 5-13 iii
T :
~ LIST OF TABLES (CONTINUED) - j Iahla Iith P.aan j i :
2 9 Effect of Isradiation to 3.28 x 10" n/cm (E > 1.0 MeV) on the Notch Toughness Properties of the CPSES Unit No. 2 Reactor Vessel Surveillance Materials . 5-14 l
i 1 5-10 Comparison of the CPSES Unit No. 2 Surveillance Material 30 ft-lb Transition l
[ Ten.perature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, !
Revision 2, Predictions 5-15 ;
l .
l 5-11 . Tensile Propenies of the CPSES Unit No. 2 Reactor Vessel Materials Irradiated to a 2
Fluence of 3.28 x 10" n/cm (E > 1.0 MeV) 5-16 6-1 Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center 6-15 I '
6-2 - Calculated Azimuthal Variation of Fast Neutron at the Pressure Vessel Clad / Base Metal Interface 6-16
]
l 6-3 ~ Relative Radial Distribution of & (E > 1.0 MeV) Within the Pressure Vessel Wall 6-17 l
6-4 Relative Radial Distribution of & (E > 0.1 MeV) Within the Pressure Vessel Wall 6-18 j 6-5 Relative Radial Distribution of dpa/sec Within the Pressure Vessel Wall 6-19 6-20 6-6 Nuclear Parameters Used in the Evaluation of Neutron Sensors 6-7. Monthly 'Ihermal Generation During the First Fuel Cycle of the CPSES Unit 2 l Reactor 6-21 i o !
{
6-8 Measured Sensor Activities and Reaction Rates, Surveillance Capsule V, Saturated 4
Activities and Reaction Rates 6-22 l 6-9 Summary of Neutron Dosimetry Results Surveillance Capsules U 6-24 !
6-10 Comparison of Measured and FERRET Calculated Reaction Rates at the j 7
Surveillance Capsule Center, Surveillance Capsule U 6-25 ;
6-11 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule U 6-26 !
6-12 Comparison of Calculated and Measured Neutron Exposure Levels for CPSES Unit 2
, Surveillance Capsule U 6-27 i i
, iv -
l
LIST OF TABLES (CONTINUED)
Iahls Iitig Ea_gs 6-13 Neutron Exposure Projections at Key Locations on the Pressure Vessel Clad / Base Metal Interface 6-28 6-14 Neutron Exposure Values for the CPSES Unit 2 Reactor Vessel 6-29 6-15 Updated Lead Factors for CPSES Unit 2 Surveillance Capsules 6-30 7-1 Recommended Surveillance Capsule Removal Schedule for the CPSES Unit No. 2 .
Reactor Vessel 7-1 y
t IJST OF ILLUSTRATIONS 1 i
fiS1EG .T111G f.R8G l 4-1' ' Arrangement of Surveillance Capsules in the CPSES Unit No. 2 Reactor Vessel 4-7--
l 4-2 Capsule U Diagram Showing location of Specimens, Thermal Monitors, and Dosimeters 4-8 i 5-1 Charpy V-Notch Impact Properties for CPSES Unit No. 2 Reactor Vessel Intermediate ;
Shell Plate R3807-2 (Iangitudinal Orientation) 5-17 !
5-2 Charpy V-Notch Impact Properties for CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807 2 (Transverse Orientation) 5-18 ;
5-3 Charpy V-Notch Impact Propenies for CPSES Unit No. 2 Reactor Vessel Surveillance W eld M etal 5-19 5-4 Charpy V-Notch Impact Propenies for CPSES Unit No. 2 Reactor Vessel Heat-Affected-Zone (HAZ) Metal 5-20 l
5-5 Charpy Impact Specimen Fracture Surfaces for CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807-2 (longitudinal Orientation) 5-21 5-6 Charpy Impact Specimen Fracture Surfaces for CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807-2 (Transverse Orientation) 5-22 5-7 Charpy Impact Specimen Fracture Surfaces of the CPSES Unit No.2 Reactor Vessel Weld Metal 5-23 5-8 Charpy Impact Specimen Fracture Surfaces of the CPSES Unit No. 2 Reactor Vessel Heat Affected-Zone (HAZ) Metal 5-24 5-9 Tensile Properties for CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807-2 (longitudinal Orientation) 5-25 5-10 Tensile Properties for CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807 2 (Transverse Orientation) 5-26 5-11 Tensile Propenies for CPSES Unit No. 2 Reactor Vessel Weld Metal 5-27 5-12 Fractured Tensile Specimens from CPSES Unit No. 2 Reactor Vcssel Intermediate Shell Plate R3807-2 (langitudinal Orientation) 5-28 Vt
LIST OF ILLUSTRATIONS (CONTINUED)
Figure Title Eage 5-13 Fractured Tensile Specimens from CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807-2 (Transverse Orientation) 5-29 5-14 Fractured Tensile Specimens from CPSES Unit No. 2 Reactor Vessel Weld Metal 5 30 I
5-15 Engineering Stress-Strain Curves for Intermediate Shell Plate R3807 2 Tensile 6
- Specimens CLI and CL2 (Longitudinal Orientation) 5-31 5-16 Engineering Stress-Strain Curve for Intermediate Shell Plate R3807-2 Tensile Specimen CL3 (Longitudinal Orientation) 5-32 5 17 Engineering Stress-Strain Curves for Intermediate Shell Plate R3807-2 Tensile Specimens CTl and CT2 (Transverse Orientation) 5-33 [
i 5-18 Engineering Stress-Strain Curve for Intermediate Shell Plate R3807-2 Tensile Specimen CT3 (Transverse Orientation) 5 34 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens CW2 and CW1 5-35 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen CW3 5-36 6-1 Plan View of a Dual Reactor Vessel' Surveillance Capsule 6-14 .
i vii
SECTION 1.0
SUMMARY
OF RESULTS
'Ihe analysis of the reactor vessel materials contained in surveillance capsule U, the first capsule to be removed from the Texas Utilities Electric Company (TU Electric) Comanch Peak Steam Electric Station (CPSES) Unit No. 2 reactor pressure vessel, led to the following conclusions:
- The capsule received an average fast neutron fluence (E > 1.0 MeV) of 3.28 x 1028n/cm' after 0.904 effective full power years (EFPY) of plant operation.
- Irradiation of the reactor vessel intermediate shell plate R3807-2 Charpy spha. oriented j with the longitudinal axis of the specimen parallel to the major rolling duection of the plate (longitudinal orientation), to 3.28 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition
- temperature shift of O'F and a 50 ft-lb transition temperature shift of O'F. This results in an irradiated 30 ft-lb transition temperature of -5*F and an irradiated 50 ft-lb transition temperature of 35'F for the longitudinally-oriented specimens.
- Irradiation of the reactor vessel intermediate shell plate R3807 2 Charpy specunen, oriented .
with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation), to 3.28 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 25 F and a 50 ft lb transition temperature increase of 20*F. This results in an irradiated 30 ft-lb transition temperature of 15'F and an irradiated 50 ft-lb transition temperature of 75*F for transversely-oriented specimens.
- Irradiation of the weld metal Charpy specimens to 3.28 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature shift of O'F and a 50 ft-lb transition temperature shift of 0 F.
This results in an irradiated 30 ft lb transition temperature of -45'F and an irradiated 50 ft-lb transition temperature of 5'F.
I l
. Irradiation of the weld Heat-Affected Zone (HAZ) metal Charpy specimens to 3.28 x 102' l n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature shift of 0'F and a 50 ft-lb transition temperature shift of 0 F. This results in an irradiated 30 ft-lb transition temperature ;
l of -105'F and an irradiated 50 ft-lb transition temperature of -50'F.
1-1
_+ ,- -- , m r +
- The average upper shelf energy of the intermediate shell plate R3807-2 (longitudinal orientation) resulted in an average energy increase of 3 ft-lb after irradiation to 3.28 x 10" n/cm2 (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 118 ft-lb for the longitudinally oriented specimen. (Since neutron irradiation is expected to decrease the upper shelf energy, this does not represent a measurable change.)
- The average upper shelf energy of the intermediate shell plate R3807-2 (transverse orientation) resulted in an average energy increase of 4 ft-lb after irradiation to 3.28 x 10" n/cm2 (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 88 ft-lb for the transversely-oriented specimens. (Since neutron irradiation is expected to decrease the upper shelf energy, this does not represent a measurable change.)
. The average upper shelf energy of the weld metal Charpy specimens resulted in an average 2
energy decrease of 9 ft lb after irradiation to 3.28 x 10" n/cm (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 85 ft-lb for the weld metal specimens.
- The average upper shelf energy of the weld ilAZ metal resulted in an average energy increase 2
of 1I ft-lb after irradiation to 3.28 x 10" n/cm (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 127 ft-lb for the weld HAZ metal. (Since neutron irradiation is expected to decrease the upper shelf energy, this does not represent a measurable change.)
. A comparison of the CPSES Unit No. 2 Surveillance Capsule U test results with the Regulatory Guide 1.99, Revision 2m, predictions led to the following conclusions:
- All measured 30 ft-lb transition temperature shift values are less than the Regulatory Guide 1.99, Revision 2, predictions (Table 5-10).
- All surveillance capsule material measured upper shelf energy decreases are less than the Regulatory Guide 1.99, Revision 2, predictions (Table 5-10).
. 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 10 CFR 50, Appendix Gm 1-2
The calculated end-of-life (32 EPPY) maximum neutron fluence (E > 1.0 MeV) for the CPSES Unit No. 2 reactor vessel is as follows:
Vessel inner radius * = 2.836 x 10 n/cm2 Vessel 1/4 thickness = 1.512 x 10 n/cm2 2
Vessel 3/4 thickness = 3.063 x 10'8 n/cm
- Clad / base metal interface i
i I
t t
9 I
I I
i 9
4 i
i t
I e
T i
4 a
J 1-3
SECTION
2.0 INTRODUCTION
This report presents the results of the examination of Capsule U, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the TU Ek,ctric CPSES Unit No. 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for CPSES Unit No. 2 was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillan program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-10684, " Texas Utilities Generating Company Comanche Peak Unit No. 2 Reactor Vessel Radiation Surveillance Program m. The surveillance program was pinnned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." Capsule U was removed from the reactor after 0.904 EFPY of exposure and shipped to the Westinghouse Science and Technology Center Hot Cell Facility, where the postirradiation machanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance capsule U removed from the TU Electric CPSES Unit No. 2 reactor vessel and discusses the analysis of the data.
I i
l 2-1
w.-s.~o -.= .s, m a..n,a-se sa.es a amm x m.,
4
- e, O
l i.
4 h-t b.
f f
1
(
1 6-1 d i I
i 8
i -
i, I
,l e
)
'4 1
( -
n i i 1
~l 3 .
s !
i I'
d f
l l
1
, I
- i. ;
4 *
+
s
_ - - . . . . . .. . - -=.. - -_
i j
SECTION 3.0 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties oflow alloy, ferritic pressure vessel steels such as A533 Grade B Class 1 (base material of the CPSES Unit No. 2 reactor pressure vessel beltline) are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure," Appendix G to Section III of the ASME Boiler j and Pressure Vessel Codek3. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTym). i RTym is defined as the greater of either the drop weight nil-ductility transition temperature (ND'IT per !
ASTM E-208") or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) :
temperature as determined from Charpy specimens oriented normal (transverse) to the major working ,
direction of the plate or forging. The RTym of a given material is used to index that material to a !
M1 reference stress intensity factor curve (Km curve) which appears in Appendix G to the ASME Code .
'Ibe K, 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 Km curve, allowable stress intensity factors can be obtained for this material as a function of temperamre.
Allowable operating limits can then be determined using these allowable stress intensity factors.
RTym 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 changes in mehanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor surveillance Dl program, such as the CPSES Unit No. 2 reactor vessel radiation surveillance program , in which a surveillance capsule is periodically removed from the opc ating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V notch 30 ft-lb temperature (ARTy m) due to hradiation is added to the initial RTym to adjust the RTym (ART) for radiation embrittlement. This j ART (initial RTm + ARTym) is used to index the material to the K, curve and, in turn, to set 3-1 l
g-
. operating limits for the melear power plant which take into account the effects of irradiation on the ' '
reactor vessel materials.
J J
.i.
d 4
}
l,
. 1
[
i 1
l
- l
-i
?
3-2 i
Ii
I SECTION
4.0 DESCRIPTION
OF PROGRAM Six surveillane capsules for monitoring the effects of neutron exposure on the CPSES Unit 2 reactor !
pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the thermal shield and the !
vessel wall as shown in Figure 4-1. 'Ihe vertical center of the capsules is opposite the vertical mater f of the core. The capsules contam specimens made from the intermediate shell plate R3807-2 and weld metal fabricated with 3/16-inch Mil B-4 weld filler wire (heat number 89833, Ilnde 124 flux, and lot number 1061) which is identical to that used in the actual fabrication of the intermediate to lower shell
- ginh weld seam of the reactor pressure vesset j
. Capsule U was removed after 0.904 effective full power years (EFPY) of plant operation. This
- capsule contained Charpy V-notch, tensile and compact fracture marhanics specimens made from l intermediate shell plate R3807-2 and submerged arc weld metal identical to the closing girth weld seam. In addition, this capsule contamed Charpy V-notch specimens from the weld Heat-Affected-Zone (HAZ) of plate R3807-2. '
Test material obtained from the intermediate shell plate (after the thermal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched ends of the plate. All test specimens were machined from the 1/4 and 3/4-thickness location of the plate after performing a l
simulated postweld, stress-relieving treatment on the test material and also from weld and heat-afi ;cd-zone metal of a stress-mlieved weldment joining intermediate shell plate R3807-2 and adjacent lower shell plate R3816-2. All beat-affected-zone specimens were obtained from the weld
]
heat-affected-zone of lower shell plate R3807 2.
2 k
- Charpy V-notch impact specimens corresponding to ASTM 370 Type A were machined from
- . intermediate shell plate R3807 2 in both the longitudinal orientation (longitudinal axis of specimen j parallel to major rolli.ig direction) and transverse orientation (longitudinal axis of specimen normal to
! major rolling direction). The mre region weld Charpy impact specimens were machined from the l weldment such that the long dimension of the Charpy specimen was normal to the weld direction.
i
'Ihe notch was machined such that the direction of crack propagation in the specimen was in the
. welding direction.
i
, 4-1
!~
--~-,vas- -
, ,r , e- , , , 4 -- - : - - - - - - - - - - - - - - - - - - - - -
The chemical composition and heat treatment of the surveillance material is presented in Tables 4-1
' through 4-4. The data in Tables 4-1 through 4-4 was obtained from Appeixlix A of the unirradiated surveillance program, WCAP-10684t'3 Capsule U contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of neptunium (Np 2") and uranium (U )23were placed in the capsule to measure the integrated flux at specific neutron energy Icvels.
The capsule contained thermal monitors made from two low-melting-point eutectic alloys and sealed in Pyrex tubes. These thermal monitors were used to defm' e the maximum temperature attained by the
' test specimens during irradiation. The composition of the two eutectic alloys and their melting points are as follows:
2.5% Ag,97.5% Pb Melting Point: 579 F (304*C) 1.5% Ag,1.0% Sn,97.5% Pb Melting Point: 590 F (310*C)
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in {
Capsule U is shown in Figure 4-2.
l i
l 1
i i
4-2
m TABLE 4-1 Chemical Composition (wt%) of the CPSES Unit No. 2 Reactor Vessel Intermediate Shell Platem Intermediate Shell Intermediate Shell Intermediate Shell Intermediate Shell Element Plate R3807-1(') Plate R3807-2('6) Plate R3807-26d Plate R3807-3)
C .21 .22 .22 .22 Mn 1.42 1.40 1.36 1.30 P .006 .007 .014 .007 S .015 .016 .014 .009 Si .25 .24 .25 .19 Ni .64 .64 .62 .60 Mo .60 .59 .58 .58 Cr .05 .04 .056 .06 Cu .06 .06 .065 .05 4 1
/d .020 .025 .018 .023 1
Co .012 .013 .014 .009 Pb <.001 <.001 .002 <.001 W <.01 <.01 <.01 <.01 Ti <.01 <.01 .0(M <.01 Zr <.001 <.001 <.002 <.001 V .002 .003 <.002 .002 Sn .003 .0(M .002 .003 As .0(M .005 .004 .005 Cb <.01 <.01 <.002 < .01 N3 .009 .010 .008 .007 B <.001 <.001 <.001 <.001 Noms:
- a. Chemical Analysis by Combustion Engineering. Inc.
- b. Surveillance program test plate.
- c. Chemical Analysis by Westinghouse, l
l 4-3 1
m -_ . _- _ _
==
TABLE 4-2 Chemical Composition (wt%) of the CPSES Unit No. 2 Reactor Vessel lower Shell Plate
- Lower Shell Plate IAwcr Shell Plate Lower Shell Plate Element R3816-1 R3816-2 R3816-3 C .23 .23 .22 Mn 1.48 .148 1.50 P .001 .002 .008 S .004 .012 .008 Si .19 .21 .19 Ni .59 .65 .63 Mo .49 .50 .52 Cr .03 .03 .04 Cu .05 .03 .04 Ad .026 .026 .018 Co .02 .012 .012 Pb <.001 <.001 <.001 W <.01 <.01 <.01 Ti <.01 <.01 <.01 Zr <.001 <.001 <.001 V .003 .003 .003 Sn .001 .001 .002 As .009 .011 .015 Cb <.01 <.01 <.01 N3 .028 .014 .014 B < .001 <.001 <.001 M
- a. Chemical Analysis by Combustion Engineering. Inc.
4-4
I TABLE 4-3 Chemical Composition (wt%) of the CPSES Unit No. 2 Reactor Vessel Weld Metal Intermediate and Lower Shell Closing Girth IAngitudinal Weld Seams Weld Seam .
Surveillance Sample Production Weldment (Identical Wire Flux Test Weld'"*) Seam No. Wire Flux Test to the Closing Girth Element Weld Sample") 101-142A Weld Sample") Seam Weld)")
C .16 .16 .088 .11 Mn 1.32 1.24 1.33 1.37 P .005 .004 .004 .011 S .011 .009 .010 .014 Si .16 .19 .51 .49 Ni .05 .08 .03 .072 Mo .54 .59 .54 .59 Cr .02 .02 .03 .058 Cu .07 .05 .05 .030 l Al -- .W -- .006 Co -- .011 -- .008 Pb -- <.001 --
.001 ,
<.01 l W -- .01 --
Ti -- <.01 --
.002 Zr -- .001 --
<.002 V .0(M .005 .003 <.002 Sn -- .003 -- .003 As -- .021 --
.018 Cb -- <.01 --
<.002 N2 -- .007 --
.008 B -- .001 --
.001 .
joTES. I
- a. Chemical Analyses by Combustion Engineering. Inc.
- b. Actual Beltline production weld chemistry (Lower Shell plate Seam No. 101 142A).
- c. Chemical Analysis by Westinghouse of the Surveillance Program Test Weldment (Test Plate "D") supplied by Combustion Engineering. Inc. (Analysis results contain an error band of +/- 10%. Standards are traceable to the National Institute of Standards & Technology and are run with each group of samples.)
l l
4-5
TABLE 4-4 Heat Treatment of the CPSES Unit No. 2 Reactor Vessel Beltline Region Surveillance Material Material Temperature ('F) Time'd (br) Cooling Austenitizing: 4 Water-quenched 1600
- 25 Intermediate (871 C)
Shell Plates Tempered: 4 Air-cooled R3807-1 'nd 1225
- 25 R3807-2 (663'C)
R3807-3 Stress Relief: 19.25N Furnace-cooled 1150 i 50 (621 C)
Austenitizing: 4 Water-Quenched 1600* 25 Lower (871 C)
Shell Plates Tempered: 4 Air-Cooled R3816-1 and 1225 25 R3816-2 (663 C)
R3816-3 Stress Relief: 14.5
- Furnace-Cooled 1150
- 50 (621 C)
Intermediate 19.25
- Furnace-cooled Shell longitudinal Stress Relief:
Seam Welds 1150 50 Imwer Shell 14.5'" Furnace-cooled longitudinal Seam Welds Intermediate to Lower Local 8.0 Furnace-cooled Shell Girth Seam Stress Relief:
Weld 1150 i 50 (621 C)
Surveillance Program Test Material Surveillance Program Weldment Test Plate "D" Post 'Neld (Representative of Stress Relief: 8.5'O Furnace-cooled closing Girth Seam) 1150 i 50 (621*C)
,40Ek
- a. Lukens Steel Company, Marrel Freres and Combustion Engineering. Inc. Certification Reports,
- b. Stress Relief includes the Intermediate to Lower Shell Closing Girth Seam Post Weld Heat Treatment.
- c. The Stress Retief Heat Treatment received by the Surveillance Test Weldment has been simulated.
4-6
O' REACTOR VESSEL CORE BARREL NEUTRON PAD (301.5 ') Z
, CAPSULE U (58.5 ')
e-
~
- 58.5
- l V (61 ')
58.5* %
61*
270' 90' (241 *) y l l (238.5') . X W (121.5')
i REACTOR VESSEL l 180*
t PLAN VIEW .
i
{ VESSEL N WALL N
- N CAPSULE s/ ASSEMBLY CORE ; _
- lltll111111 : i,(s
- ; N - CORE
{. -
- N MIOPLANE
, ; i N e%
- s\ NEUTRON PAD
- s. . . . . . . . . . . ,,
g i :%
/ CORE BARREL ELEVATION VIEW l
l Figure 4-1 Arrangement of Surveillance Capsules in the CPSES Unit No. 2 Reactor Vessel l
l 4-7 I
e -
W t
i' LEGEND: CL INTERMEDIATE SHELL PLATE R3807-2 (LONGITUDINAL)
CT INTERMEDIATE SHELL PLATE R3807-2 (TRANSVERSE)
CW WELD METAL CH - HEAT-AFFECTED-ZONE MATERIAL unos SPActRS TDessLas CCtfPACTS CoelPACTS C4tAAPTS CMAAPTS CMARPTS CORIPACTS COIIPACTs CNAAPYS CHAAPTS
== '"' '"" '"" " " " " * * "' '
Tcx CW2 CW4 CW3 CW2 CW1 CW14 Cn14 CW11 CH11 CWB CHS CL4 CL3 CL2 CLI CW5 CHS CW2 >
u CW1 CW13 C813 OWIO CHIO CW7 CH7 CW4 CH4 CW1 h
-..q n dl Cu - a g gg ? Al .15%Co Cu r: rrIgd, e ll 1 1I II l gl!
Fe !l 'I I Fe I 'I udu u u'Ia 579'F 8' ' 1 590'F r"t r 1 f0NITOR I I ll lF- Al .15%Co (Cd)
MONITOR a Il ll.
I I8 l ll 1:
I I II l "' g 11 ffU l .! 1: i i !! !Y TO TOP OF VESSEL T
CENTER Fq
_ _ - ~
[N) $Y{2[O APERTUd CARD Also Avh!!abla on Ap3rture Card v
i i
88888019115 19mam esturve cesanpfe genaarve enIMPTS CelafrTS cesspacTS CoelpacTS TuleMS CL3 CTIS CLil Cit! CL12 CT9 CLS CT8 CLS CT3 CL3 CT3 I
913 CL2 CT14 CL14 CT11 CLit cts CL5 CT5 CL5 CT2 CL2 CT4 CT3 Of2 CT1 CT2 h
_ _ _ 3 CL1 CT13 CLt3 CTio CLIO CTF CL7 CT4 Cl4 CT1 CLt CTI h
l l
l Al .15%Co Cu 8 ll l Al .15%Co !
l 18 i re i uLlIUI A1 15%Co (Cd) 57 F
, Al .15%Co (Cd) ;
I ll ll 1 i Hi l! 1.ll 11 l[ 1l Ni '
1 l
TO BOTTOM OF VESSEL ION OF VESSEL 3 l
Figure 4-2 Capsule U Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters h b$ _
... l SPETION 3.0 TESTING OF SPECIMENS PROM CAPSULE U ';
i 5.1 Overview ,
'Ibe post ~u radiatian mechanical testing of the Charpy V-notch impact spacimans and tantile spei-e ~ i was performed in the Remote Men agraphic Facility at the Wehe -;* Science and Tehnnlogy Center. . Testing was performed in accordance with 10 CPR 50,4radiv If'1, ASTM Specification i E185-82U3 and Westinghouse Procedure MHL 8402, Revision 2 as modified by Wati==hanse RMF l Procedum 8102, Revision 1, and 8103, Revision 1. j i
Upon receipt of the capsule at the hot celllaboratory, the specunens and spacer blocks were carefully removed, iama-*~I for identification number, and checked against the master list in WCAP-10684m, .l 1
No dism=9~ were found. j Examination of the two low-melting point 579'F (304*C) and 590'F (310'C) eutectic alloys indicated -
l I
no melting of either type of thennal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579 F (304'C). ;
i
'the Charpy impact tests were performed per ASTM Specification E23-93a* and RMF :
. Procedum 8103, Revision 1, on a Tinius-Olsen Model 74,358J machine. 'Ihe tup (striker) of the !
Charpy impact test mehine is ~ instrumented with a GRC 830-I instmmentation system, feeding j i information into an IBM compatible 486 computer. With this system, load-time and energy-time !
t signals can be mcorded in addition to the standard maamrement of Charpy energy (FJ. From the ;
j load-time curve (Appendix A), the load of general yielding (Pw), the time to general yielding (tm),
the maximum load (Pu), and the, time to maximum load (tu) can be determined. Under some test i conditions, a sharp drop in load inoicative of fast fracture was observed. The load at which fast
- fracare was initiated is identified as the fast fracture load (P,), and the load at which fast fracture terminated is identified as the anest load (P4 ).
f}
'Ihe energy at maximum load (E )u was determined by companng the energy-time record and the load-time record. . 'Ihe energy at maximum load is approximately equivalent to the energy required to L . initiate a crack in the specimen. Therefore, the propagation energy for the crack (E,) is the diffemace
' between the total energy to fracture (FJ and the energy at maximum load (Eu ).
b 5-1
The yield stress (o y) was calculated from the three-point bend formula having the following expressioc:
oy= (Pay
- L) / [B * (W - a)2
- CJ (1) where: L
~
= distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W. -= height of the specimen, measured perpendicularly to the notch a = notet depth The constant C is dependent on the notch flank angle ($), notch root radius (p) and the type of loading (ie. pure bending or three-point bending). In three-point bending, for a Charpy specimen in which & = 45' and p = 0.010 inches, Equation 1 is valid with C = 1.21. Therefore (for L = 4W),
oy= (P ay
- L) / [B * (W - a)2
- 1.21] = (3.33
- P ay
- W) / [B * (W - a)2] (2)
For the Charpy specimen, B = 0.394 inches, W = 0.394 inches and a = 0.079 inches, Equation 2 then reduces to:
oy= 33.3
- P ay (3) where o ris in units of psi and Pay is in units of pound. 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 through 5-7 is the cross-sectional 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"). 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 lastron, split-console test machine (Model 1115) per ASTM Specification E8-9305 and E21-92["3, and RMF Procedure 8102, Revision 1. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.
5-2
. _ _ _ __ _ . . _ . _ ._ -. __ _ _ y ..
!- l l r Extension measurements were made with a lmear variable displacement transducer extensometer. The i' extensometer Isife edges wem spring-loaded to the guimen and operated through spaciman failure. !
1
'Ibe extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASIM E83-93"23 i I Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. !
(
Because of the difficulty'in remotely attaching a thermocouple directly to the specimen, the following l pmcedure was used to monitor specimen temperatums. Chromel-alumel thermocouples were positioned at center and each end of the gage section of a dummy specunen and in each grip. In the test configuration, with a slight load on the spaciman. a plot of specimen temperature versus upper and
.f L lower grip and controller temperatures was developed over the range from room temperature to 550'F - i (288'C). The upper grip was used to control the furnace temperature. During the actual testing, the -j grip temperatures were used to obtain desired specimen temperatures. Experiments indicate that this -
(
method is accurate to i 2*F. i i
I i
The yield load, ultimate load, fracture load, total elongation, and uniform elongation wem determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. 'Ibe final diameter and final gage length were determined from post-fracture photographs. 'Ibe fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5.2 Charov V-Nntch imnact Test Raculte
'Ihe results of the Charpy V-notch impact tests performed on the various materials contained in i Capsule U, which was irradiated to 3.28 x 10" n/cm 2(E > 1.0 MeV), are presented in Tables 5-1 D3 through 5-8 and are compared with unirradiated results as shown in Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule U surveillance
, . materials are summarized in Table 5-9.
i Irradiadon of the reactor vessel intermediate shell plate R3807-2 Charpy specimens oriented with the .
l longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal oriantation) to 3.28 x 10" n/cm2 (E > 1.0 MeV) (Figure 5-1) resulted in a 30 ft-lb transition temperature shift of O'F and a 50 ft-lb transition temperature shift of 0 F. 'Ihis results in an irradiated 5-3
30 ft-lb transition temperature of-5'F and an irradiated 50 ft lb transition temperature of 35 F {
- (longitudinal orientation). !
I The average upper shelf energy (USE) of the intermediate shell plate R3807-2 (longitudinal {
l 2
orientation) resulted in an energy increase of 3 fi lb after irradiation to 3.28 x 10 n/cm .j (E > 1.0 MeV). This results in an irradiated average USE of 118 ft-lb (Figure 5-1).
i i'
Irradiation of the reactor vessel intermediate shell plate R3807-2 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse
. orientation) to 3.28 x 10 n/cm2 (E > 1.0 MeV) (Figum 5-2) resulted in a 30 ft-lb transition f temperature increase of 25'F and a 50 ft-lb transition temperature increase of 20*F. This results in an irradiated 30 ft-lb transition temperature of 15'F and an irradiated 50 ft-lb transition temperature of i
- 75'F (transverse orientation). (The notch in the Charpy specimen of the " transverse" specimen is !
orientated such that the direction of crack propagation is in the rolling direction.) {
i
'Ihe average upper shelf energy (USE) of the intermediate shell plate R3807-2 Charpy specimens (transverse orientation) resulted in an energy increase of 4 ft lb after irradiation to 3.28 x 10 n/cm2 j (E > 1.0 MeV). This results in an irradiated average USE of 88 ft-lb (Figure 5-2). f l.
2 Irradiation of the surveillance weld metal Charpy specimens to 3.28 x 10** n/cm (E > 1.0 McV) l 6
(Figure 5-3) resulted in a 30 ft-lb transition temperature shift of 0'F and a 50 ft-lb transition ,
temperature shift of O'F. This results in an irradiated 30 ft-lb transition temperature of-45'F and an f irradiated 50 ft-lb transition temperature of 5'F. !
The average USE of the surveillance weld metal resulted in an energy decrease of 9 ft-lb after i irradiation to 3.28 x 10 n/cm 2(E > 1.0 MeV). This results in an irradiated average USE of i
85 ft-lb (Figure 5-3). !
I 2
Irradiation of the reactor vessel HAZ metal Charpy specimens to 3.28 x 10 n/cm (E > 1.0 MeV) l (Figure 5-4) msulted in a 30 ft-lb transition temperature change of 0 F and a 50 ft-lb transition ;
J temperature change of 0*F. This results in an irradiated 30 ft lb transition temperature of -105 F and j an irradiated 50 ft-lb transition temperature of -50*F. l The average USE of the weld HAZ metal resulted in an energy increase of 11 ft-lb after irradiation to 3.28 x 10 n/cm 2(E > 1.0 MeV). 'Ihis results in an irradiated average USE of 127 ft-lb (Figure 5-4). !
l l
5-4 ,
i
'Ihe fracture appearance of each irradiated Cha py specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test temperature.
' A comparison of the CPSES Unit No. 2 reactor vessel beltline material test results with the Regulatory Guide 1.99, Revision 2, predictions (Table 5-10) led to the following conclusions:
-- All measured 30 ft-lb transition temperature shift values of the surveillance materials are less than the Regulatory Guide 1.99, Revision 2, predictions.
- All measured upper shelf energy (USE) percent decreases are less than the Regulatory Guide 1.99, Revision 2, predictions.-
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 10 CPR 50, Appendix Gr2i, The load-time records for individual instrumented Charpy specimen tests are shown in Appendix A.
5.3 Tensile Test Results (
i !
j The results of the tensile tests performed on the various materials contained in Capsule U irradiated to l
2 l
3.28 x 10" n/cm (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated .
i resultst 'l as shown in Figures 5-9 through 5-11. I i
i The results of the tensile tests performed on the intermediate shell plate R3807-2 (longitudinal 2
orientation) indicated that irradiation to 3.28 x 10" n/cm (E > 1.0 MeV) caused a 2 to 7 ksi inaease I
j in the 0.2 percent offset yield strength and a 3 to 5 ksi increase in the ultimate tensile strength when t
compared to unirradiated data') (Figure 5-9).
4
'lhe results of the tensile tests performed on the intermediate shell plate R3807-2 (transverse 2
orientation) indicated that irradiation to 3.28 x 10" n/cm (E > 1.0 MeV) caused a 2 to 12 ksi increase in the 0.2 percent offset yield strength and a 3 to 5 ksi increase in the ultimate tensile strength when compared to unirradiated datat ') (Figure 5-10).
5-5
The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 3.28 x 10" n/cm (E2
> 1.0 MeV) caused a 3 to 7 ksi increase in the 0.2 percent offset yield strength - _
D1 and a 5 to 9 ksi increase in the ultimate tensile strength when compared to unirradiated data (Figure 5-11).
The fractured tensile specimens for the intermediate shell plate R3807-2 material are shown in Figures 5-12 and 5-13, while the fractured specimens for the surveillance weld metal are shown in Figure 5-14.
The engineering stress-strain curves for the tensile tests are shown in Figures 5-15 through 5-20.
5-6
I TABLE 5-1 Chagy V-Notch Impact Data for the CPSES Unit No. 2 Intermediate Shell Plate 2
R3807-2 Irradiated to a Fluence of 3.28 x 10" n/cm (E > 1.0 MeV)
(IAngitudinal Orientation)
Sample Temperature impact Energy Lateral Expansion Shear Number ( F) ( C) (ft-lb) (J) (mils) (mm) (%)
CL2 -75 -59 8 11 8 0.20 2 CL11 -50 -46 11 15 10 0.25 5 CLIO -25 -32 20 27 17 0.43 10 CLl3 -10 -23 36 49 27 0.69 15 CL9 0 -18 41 56 30 0.76 15 CL5 10 -12 42 57 32 0.81 20 CL15 25 -4 42 57 32 0.81 25 CL3 50 10 58 79 46 1.17 35 CL4 72 22 85 115 61 1.55 40 CLl4 100 38 88 119 65 1.65 45 CL6 125 52 106 144 71 1.80 60 CL7 150 66 115 156 80 2.03 100 CL8 200 93 116 157 81 2.06 100 CL1 250 121 116 157 77 1.% 100 CL12 300 149 127 172 84 2.13 100 57
1 i
TABLE 5-2 Charpy V-Notch Impact Data for the CPSES Unit No. 2 Intermediate Shell Plate 2
R3807-2 Irradiated to a Fluence of 3.28 x 10" n/cm (E > 1.0 MeV)
(Transverse Orientation)
Sample Temperature Impact Energy Lateral Expansion Shear Number ( F) ( C) (ft-lb) (J) (mils) (mm) (%) i CT3 -125 -87 6 8 6 0.15 0 CT5 -95 -71 4 5 5 0.13 0 CT12 -50 -46 20 27 14 0.36 5 CT9 -25 -32 22 30 16 0.41 10 CT15 -10 -23 29 39 24 0.61 10 CT2 0 -18 24 33 22 0.56 15 CTI 10 -12 26 35 20 0.51 15 CTI1 50 10 35 47 35 0.89 20 CT13 72 22 66 89 50 1.27 30 CT10 100 38 48 65 46 1.17 25 CT7 125 52 73 99 54 1.37 30 CT8 150 66 85 115 67 1.70 100 CT6 200 93 87 118 70 1.78 100 CT14 250 121 86 117 67 1.70 100 CT4 300 149 % 130 69 1.75 100 5-8
1 TABLE 5-3 Charpy V-Notch Impact Data for the CPSES Unit No. 2 Surveillance Weld Metal 2
Irradiated to a Fluence of 3.28 x 10 8 n/cm (E > 1.0 MeV)
Sample Temperature Impact Energy Lateral Expansion Shear Number ( F) ( C) (ft-lb) (J) (mils) (mm) (%)
CWil -125 -87 3 4 8 0.20 5 CW5 -95 -71 11 15 9 0.23 5 CW15 -75 -59 7 9 9 0.23 10 CW1 -60 -51 32 43 24 0.61 10 CW4 -50 -46 38 52 29 0.74 15 CW3 -25 -32 40 54 31 0.79 15 CW9 -10 -23 48 65 38 0.97 25 CW2 0 -18 48 65 39 0.99 35 CW7 50 10 66 89 52 1.32 70 CW14 72 22 75 102 61 1.55 85 CW6 100 38 76 103 62 1.57 90 CW12 150 66 80 108 65 1.65 100 CW10 200 93 87 118 67 1.70 100 CW8 250 121 86 117 60 1.52 100 CW13 300 149 86 117 70 1.78 100 5-9
TABLE 5-4 Charpy V-Notch Impact Data for the CPSES Unit No. 2 Heat-Affected-Zone (HAZ) 8 2 Metal Irradiated to a Fluence of 3.28 x 10 n/cm (E > 1.0 MeV)
Sample Temperature Impact Energy Lateral Expansion Shear Number ( F) ( C) (ft-lb) (J) (mils) (mm) (%)
CH9 -225 -143 7 9 2 0.05 5 CH8 -175 -115 16 22 8 0.20 5 CH3 -150 -101 20 27 9 0.23 10 CH2 -125 -87 40 54 21 0.53 15 Cll7 -100 -73 23 31 13 0.33 10 CH5 -75 -59 90 122 44 1.12 50 CIll0 -60 -51 59 80 37 0.94 30 CHil -50 -46 35 47 26 0.66 30 CH4 -25 -32 52 71 29 0.74 60 CH12 0 -18 78 106 50 1.27 65 CIll4 25 -4 90 122 59 1.50 65 CHIS 72 22 124 168 75 1.91 100 Cill3 150 66 104 141 71 1.80 100 CIll 200 93 154 209 74 1.88 100 CH6 200 93
- Specimen Alignment Error. Data is not valid.
5-10
TABLE 5-5 Instrumented Charpy Impact Test Results for the CPSES Unit No. 2 Intermediate Shell Plate R3807-2 Irradiated to a Fluence of 3.28 X 10" n/cm2 (E > 1.0 MeV) (Longitudinal Orientation)
Normalized Energies (ft-lb/in')
Time Tune Fast Charpy Yield to Max. to Fract. Arrest Yield Test Energy load Yield Load Max. Ioad load Stress Flow Sample Temp. F, Charpy Max. Prop. Pay toy Pu tu P, PA oy Stress No. ('F) (ft-lb) F4 /A Eu/A E,/A (Ib) (psec) (Ib) (psec) (Ib) (Ib) (ksi) (ksi)
CL2 -75 8 64 37 27 3737 0.16 3737 0.16 3737 144 124 124 CL11 -50 11 89 46 42 3627 0.15 3808 0.17 3808 157 120 123 CLIO -25 20 161 120 41 3455 0.14 4209 031 4209 364 115 127 3 CL13 -10 36 290 242 47 3453 0.14 4520 0.54 4520 270 115 132 b CL9 0 41 330 259 71 3454 0.14 4500 0.56 4486 229 115 132 CL5 10 42 338 292 46 3348 0.14 4534 0.63 4534 362 111 131 CLIS 25 42 338 253 85 3250 0.14 4360 0.57 4360 725 108 126 CL3 50 58 467 313 154 3204 0.14 4392 0.69 4345 1457 106 126 CL4 72 85 684 319 365 3198 0.14 4441 0.69 3740 1668 106 127 CLl4 100 88 709 310 398 3132 0.15 4324 0.69 3737 2019 104 124 CL6 125 106 854 305 549 3024 0.14 4323 0.69 3270 2052 100 122 CL7 150 115 926 303 623 2995 0.14 4218 0.69 N/A N/A 99 120 CL8 200 116 934 295 639 2840 0.14 4124 0.69 N/A N/A 94 116 l CLI 250 116 934 289 645 2741 0.14 4026 0.7 N/A N/A 91 112 CL12 300 127 1023 287 736 2703 0.14 3950 0.7 N/A N/A 90 110 N/A - Fully ductile fracture. No anest load.
. .= =.. - -. - . .. . . . _ . .- _ ._ -. _ - _ . - - _ _ _ _ _ _
l l
l l
TABLE 5-6 Instmmented Charpy Impact Test Results for the CPSES Unit No. 2 Intermediate Shell Plate R3807-2 Irradiated to a Fluence of 3.28 x 10" n/cm' (E > 1.0 MeV) (Transverse Orientation)
Normalized Energies (ft-lb/in2 )
l Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Test Energy Imad Yield lead Max. Ioad load Stress Flow Sample Temp. F, Charpy Max. Prop. Por toy Pu tu P, P, oy Stress E,/A (Ib) (psec) (Ib) (psec) (Ib) (Ib) (ksi) (ksi)
No. (*F) (ft-Ib) FJA Eu/A l
CT3 -125 6 48 29 20 3511 0.14 3511 0.14 3511 106 117 117 Cr5 -95 4 32 16 16 2351 0.11 2351 0.11 2351 79 78 78 Crl2 -50 20 161 116 45 3861 0.17 4242 031 4242 375 128 135 CP9 -25 22 177 125 52 3537 0.14 4244 032 4244 237 117 129 Crl5 -10 29 234 189 45 3442 0.14 4368 0.44 4368 366 114 130 CT2 0 24 193 130 63 3431 0.14 4129 034 4129 489 114 126 CTI 10 25 209 141 69 3429 0.16 4159 036 4159 488 114 126 Cril 50 35 282 163 119 3165 0.14 4081 0.42 4081 1053 105 120 CT13 72 66 531 305 227 3229 0.16 4348 0.67 4152 1575 107 126 Cr10 100 48 387 216 170 3096 0.14 4130 0.52 4130 2045 103 120 Cr7 125 73 588 293 295 3090 0.14 4252 0.67 3917 2403 103 122 Cr8 150 85 684 281 403 3045 0.14 4186 0.65 N/A N/A 101 120 Cf6 200 87 701 278 422 2932 0.14 4040 0.66 N/A N/A 97 116 Crl4 250 86 692 265 427 2715 0.14 3904 0.66 N/A N/A 90 110 CT4 300 % 773 273 500 2672 0.14 3919 0.67 N/A N/A 89 109 i
j N/A - Fully ductile failure. No anest load.
l
TABLE 5-7 Instrumented Charpy Impact Test Results for the CPSES Unit No. 2 Smveillance Weld hietal 2
Irradiated to a Fluence of 3.28 x 10" n/cm (E > 1.0 MeV)
Normalized Energies 2
(ft-Ib/in )
Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Load Yield Load Max. Load Load Stress Stress Sample Temp. En Charpy Max. Prop. Pay toy Pu tu P, P, oy (ksi)
No. ( F) (ft-lb) En/A Eu/A E,lA (Ib) (psec) (Ib) (psec) (Ib) (Ib) (ksi)
CWil -125 3 24 8 17 1353 0.09 1353 0.09 1353 86 45 45
('J/5 -95 11 89 53 35 3816 0.15 4070 0.19 4070 137 127 131-CW15 -75 7 56 23 33 2958 0.13 2958 0.13 2958 126 98 98 CW1 -60 32 258 219 39 3716 0.17 4399 0.5 4399 246 123 135 C; CW4 -50 38 306 247 59 3541 0.14 4434 0.54 4414 113 118 132 CW3 -25 40 322 230 92 3493 0.14 4375 031 4268 579 116 131 CW9 -10 48 387 233 154 3479 0.14 4366 0.52 4150 442 116 130 CW2 0 48 387 234 152 3419 0.14 4318 033 4113 592 114 128 CW7 50 66 531 237 294 3160 0.14 4161 036 3214 1597 105 122 CW14 72 75 604 293 311 3109 0.15 4156 0.67 3160 2100 103 121 CW6 100 76 612 234 378 3116 0.14 4082 036 2464 1679 103 120 CW12 150 80 644 281 363 2970 0.14 3980 0.67 N/A N/A 99 115 CW10 200 87 701 278 423 2817 0.14 3865 0.69 N/A N/A 94 111 CW8 250 86 692 272 421 2756 0.14 3828 0.67 N/A N/A 92 109 CW13 300 86 692 268 425 2675 0.14 3715 0.69 N/A N/A 89 106 N/A - Fully ductile failure. No arrest load.
TABLE 5-8 Instrumented Charpy Impact Test Results for the CPSES Unit No. 2 Heat-Affected-Zone (HAZ) Metal 2
Irradiated to a Fluence of 3.28 x 10 n/cm (E > 1.0 hieV)
Normalized Energies 2
(ft-lb/in )
Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Test Energy Load Yield Load Max. Ioad Load Stress Flow Sample Temp. Fo Charpy Max. Prop. Poy toy Pu tu P, P, oy Stress No. (*F) (ft-lb) Er/A EJA E,./A (Ib) (psec) (Ib) (psec) (Ib) (Ib) (ksi) (ksi)
OI9 -225 7 56 33 23 4082 0.14 4082 0.14 4082 127 136 136 CH3 -175 16 129 87 42 5211 033 5211 033 5211 89 173 173 013 -150 20 161 107 54 4642 0.18 4778 0.26 4778 191 154 156 CII2 -125 40 322 270 52 4399 0.18 5183 0.53 5183 305 146 159 k CIl7 -100 23 185 152 33 4400 0.19 4849 035 4849 99 146 154 C15 -75 90 725 '348 376 4186 0.18 4910 0.67 4114 1670 139 151 0 110 -60 59 475 261 214 4056 0.17 4900 0.54 4699 1569 135 149 CIlli -50 35 282 180 102 4087 0.18 4621 0.41 4621 1599 136 145 CIl4 -25 52 419 205 214 4099 0 17 4593 0.44 4297 2064 136 144 CII12 0 78 628 243 385 3614 0.14 4537 0.52 2902 1976 120 135 CIll4 25 90 725 258 467 3486 0.15 4464 0.56 3622 2519 116 132 CIll5 72 124 998 328 671 3461 0.15 4614 0.69 N/A N/A 115 134
- CIll3 150 tot 837 304 533 3214 0.14 4292 0.67 N/A N/A 107 125 CIII 200 154 1240 395 845 3093 0.14 4355 0.86 N/A N/A 103 124 CII6 200 * * * * * * * * * * * *
- : Specimen Alignment Error. Data is not valid. N/A - Fully ductile failure. No anest load.
TABLE 5-9 2
Effect of Irradiation to 3.28 x 10" n/cm (E > 1.0 MeV) on the Notch Toughness Properties of the CPSES Unit No. 2 Reactor Vessel Surveillance Materials Average 30 (ft-lb)
- Average 35 mil Lateral" Average 50 ft-lb N Average Energy Absorption "
Transition Temperature ( F) Expansion Temperature ("F) Transition Temperature ("F) at Full Shear (USE)(ft-Ib)
Material Unirrad. Irradiated AT Unitrad. Irradiated AT Unirrad. Irradiated AT Unirrad. Irradiated A Intermediate Shell Plate -5 -5 0 30 30 0 35 35 0 115 118 +3 R3807-2 (Longitudinal) u
- 2. Intermediate Shell Plate -10 15 25 45 45 0 55 75 20 84 88 +4 R3807-2 (Transverse)
Weld Metal -45 -45 0 -5 -5 0 5 5 0 94 85 -9 IIAZ Metal -105 -105 0 -35 -35 0 -50 -50 0 116 127 +11 (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 CPSES Unit No. 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2 Predictions 30 ft-lb Transition Upper Shelf Energy Fluence Temperature Shift Decrease (n/cm2, Material Capsule Predicted
- Measured Predicted
- Measured E > 1.0 MeV)
( F) ( F) (%) (%)
Intermediate U 3.28 x 10'8 26.9 0 -14.5 +2.6 Shell Plate R3807-2 (Longitudinal)
Intermediate U 3.28 x 1028 26.9 25 -14.5 +4.8 Shell Plate R3807-2 (Transverse)
W eld U 3.28 x 10 8 20.0 0 -14.5 -9.6 Metal IIAZ U 3.28 x 10 t8 -
0 - +9.5 Metal
.iRI!L (a) Based on Regulatory Guide 1.99 Revision 2 methodology using Mean wt. % values of Cu and Ni 5-16 i
l l
l l
TABLE 5-11 2
Tensile Properties of the CPSES Unit No. 2 Reactor Vessel Materials Irradiated to a Fluence of 3.28 x 10 n/cm (E > 1.0 MeV)
Test 0.2% Yield Ultimate Fractum Fracture Fractum Uniform Total Reduction Sample Temp. Strength Stmngth load Stress Strength Bongation Bongation in Area Material Number (*F) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) (%)
Intermediate CLI 70 71.3 91.7 3.00 184.5 61.1 113 25.2 67 Shell Plate CL2 150 68.8 88.2 2.78 154.8 56.6 10.5 24.2 63 R3807-2 (Longitudinal) CL3 550 62.4 89.0 2.90 166.8 59.1 11.3 24.2 65 Intermediate Cri 25 72.8 %.8 3.40 152.8 69.3 12.8 25.2 55 u Shell Plate CTS 150 73.6 88.6 3.07 155.5 62.5 12.0 24.8 60 b R3807-2 Uransmse) Cr3 550 61.9 89.0 3.25 175.3 66.2 11.3 21.9 62 Weld Metal CW2 25 75.9 92.7 3.04 163.9 61.9 13.5 29.3 62 CWI 150 73.3 87.6 2.85 169.4 58.1 12.0 29.6 66 CW3 550 66.2 87.6 3.15 175.4 64.2 11.3 22.8 63
l Curve 784517 Temperaturo ( C)
-100 -50 0 50 100 150 200 250 l I I I I I I I I 3 3 2 100 -
0 0 ; >-00--
- 80 -
6.
M 60 -
OO 2 O (n 40 -
20 0 20 -
0 2 100 2.5 g 80 - ;
f*b -
2.0 5 60 -
d -
E y 40 - O 8 ~
1.5i ~
-40
~
~
O O 20 -
O -
0.5 0 ( , '2 0
240 O Unirradiated 160 -
220
-
- Irradiated
~
140 -
180 120 -
O - O a A+3 -
160 4 Unirradiated D U 140 E100 -
\ 2
~
80 -
O* 100 E O W
60 - b -
80 1 40 - No0 Irradiated to -
60 3.28 x 1018 n/cm2 -
40 20 -
2\ A0 -
20 0 ' ' ' ' ' '
-200 -100 0 100 200 300 400 500 Temperature ( F)
Figure 5-1 Charpy V-Notch Impact Propenies for CPSES Unit No. 2 Reactor Vessel Intermediate ;
Shell Plate R3807 2 (Longitudinal Orientation) 5-18 l
J
Temperatura ( C)
C" *55 l
-100 -50 0 50 100 150 200 250 I i i i 3
i 3
i 2
i r 100 - O -
N g 80
~
la 60 -
i
.8 l w 40 -
e e 20 -
0 *2 100 2.5 g 80 -
2 -
2.0 A 60 -
1.5 -
E o.
W 40 - -
1.0 E l C g 3 20 - A0 . 0,5 g ,
0 - 0 240 O Unirradiated 160 -e irradiated 220
~
140 -
180 120 -
160 e ~
$.100 -
O e g; 80 Unirradiated R- -
1 A+4 -
120 S e o E
100 W
60 - -
80 0 ,
40 - 2 60 A 20 Irradiated to ~
20 -
\ A25 3.28 x 1018 n/cm2 _
0 8 ' ' ' ' I
-200 -100 0 100 200 300 400 500 Temperature ( F)
Figure 5-2 Charpy V-Notch Impact Propenies for CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807-2 (Transverse Orientation) 5-19
l l
Curve 78451S Tamperature ( C)
-100 -50 0 50 100 150 200 250 i l I i i i i i 2 3 2 100 -
0 N
- 80 - O k 60 -
2 E
w 40 -
l 20 -
\
0 100 2.5 g 80
=
o - n V
6 -
2.0
_ . a 5c. 60 -
O e -
1.5 E x O W 40 - -
1.0 E C
,8 A0 3 20 - -
0.5 0 -
0 240 O Unirradiated 160 - e Irradiated 220 200 140 -
180 l 120 -
160 l
@ Unirradiated l E.100 2
n a
^
( -
140 120 2 80 -
- 69 .
O # 100 l
W '
60 -
0 0 -
80
-2 -
60 40 -
e 0 Irradiated to O o 3.28 x 1018 n/cm2 40 20 -
20 0 * ' ' ' ' '
-200 -100 0 100 200 300 400 500 Temperature ( F)
Figure 5-3 Charpy V-Notch Impact Propenies for CPSES Unit No. 2 Reactor Vessel Weld Metal 5-20
Curve 784514 A PMS 25 Tamperaturo (*C) _
-150 -100 -50 0 50 100 150 200 1 I i 3' 2'3 I I '
100 -
O g 80 -
5 60 -
E 0
- 0 v) 40 -
o O 20 -
0 100 2.5
.0 g 80 - -
O
- k -e 5 60 -
g O -
1.5 g is 40 - # -
1.0 5 c A0 3 20 -
e -
0.5 0 0 240 0 Unirradiated 160 220
-e Irradiated e -
200 140 -
~
O O g 120 ru o A + 11 -
160 d100 - 4 o -
140 O
g e -
120 3
- 80 -
Unirradiated f 60 -
\ \lrradiated to 3.23 x 10 18/ cm2 100 80 NA0 60 40 -
e 40 20 - # 0 g -
20 0 ' ' ' ' ' '
-300 -200 -100 0 100 200 300 400 Temperature ( F)
I Figure 5-4 Charpy V Notch Impact Propenies for CPSES Unit No. 2 Reactor Vessel Heat-Affected-Zone GIAZ) Metal 5-21
'Eh.w-v .
7
~$
&[%5Y Y $$
b?@$f$
7, Nf $,[7.
23 wdJ ff i- 'y . ;
.:.x .yy, }-
f
, n Mo '- q?;; $g,-
s, . i?6' 93 ~ f::oQ. 3 s%y*,;>d E:
~
$c&+,' ' ' *. ng ,
i r#4!
a , qu%,,
f.y,,Qj;$$ l[;s
- y. -y .;y{:}:s g.
a CL2 CLil CLIO CLl3 ~Y~
i
~
}' ,us m ."- :g$m/ ~3;c.
ewp Q % "5 y [1 i r y1;p; pig ge; Iw :.9w
'y'
.s; :.y9p r.c y$
f 35~j j g .x.4[fw.p 4 .,
p (
I"'M 9
- L .;,;< : ~- y~~~ ,;& i
- g 4 ., ~
- i . . ,
h
-; /N
^
IN bI ,
't ;?.M,h % k)e,. .g , $d. w : -; ,'
"m -)p w.-;q$ -
$b$%0{A . $~.>, :; kk $;$'fNf lf, Y;- '
~
} h.l CL5 CL15 CL3 CL4 CL14 k
m p.m w p, , ,
usy s+am
- ; c. -, y_. .,.
\
3
? P CL6 CL7 CL8 CLI CL12 Figure 5 5 Charpy Impact Specimen Fracture Surfaces of the CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807-2 (Longitudinal Orientation) l 5-22
'je k,[;+
D: hs;(psiikO[
.. ;Qpy%.
[
L '$g$[ 9
- [
- M[5 f,.yt; %{ l A-; ,l.
- dis. .b \ N AN5!
ME CT3
$PJi! b# hil@ l CT5 CT12 CT9 CT15 Y
- r 1 vl .k < -
CT2 CTI CTil CT13 CTIO t
s.; p
- p a . ;,ag i
xhk f(H%t
.gT f g ',-
19 h5TEMh 3
L 3
,1 y- :' }
a i .
.k p}kkf u
pa ,
Y 5 b$fkl% _ m ?g?! $ ", j-lN 'N$
CT7 CT8 CT6 CT14 CT4 Figure 5 6 Charpy Impact Specimen Fracture Surfaces of the CPSES Unit No. 2 Reactor Vessel Intennediate Shell Plate R3807-2 (Transverse Orientation) 5-23
eg-gA( ~ '" J > .L
- .&: ;.zw R:_;L.,igt8 ,
j k'Q~
- '~ ~ , ,.
we el __
m, x W55V{t 3
- .W hhN?j I h- -
%M CW11 T R$ g %%
CW5
@ CW15 CW1 la CW4 7.c :
.-,& ut ,, :?
9
.] S4
- ~ .
5 MRh, ..l ,
- . Gj *
~k !
hk$h h I5ue.'b Y(Y$" 5:
CW3 CW9 CW2 CW7 CW14 Y-
, 9g g r, . ,, .y; n x;- n. m jl.$ &, p *., n 4)5;il j:ST%_ +- $}@Qill ca ! .
~ m tQec
.~. :, -
'i aW{
q e:9f '
T' A< g a
%jym(p
'- J:m p,y;q p;n 2.sccp .~ a
.s-c
.e@g, s, . ,n CW6 CW12 CW10 CWS CW13 Figure 5-7 Charpy Impact Specimen Fracture Surfaces of the CPSES Unit No. 2 Reactor Vessel Weld Metal 5-24
b 'h)
- h. : .{'?.]$'. W:}fh,b GG;?:<
3tus
> g):;/c..y s.9 )':e 1 w: . a.
9.p"l .L&t A.-d N
l MjfM Y 5.; 'y [ .. s.]y$ :;~5v
,- y%;
- 3. .i);ll b<4y ;jd,i;,
>ge n/;* w:n ; '. ~9 (th gn , h;g q;(NO;:.,
+ L.
n V :-
W%m dj[.:N f '
+*4).4 iYB
- 4mgj}
- t ,f Tgl%?;
- 1. m 3 ; y CH9 CHS CH3 CH2 CH7 I,
- -)
9
.,4 c.:n wka R -
4
- [ E,(dh.[.,$. {kM[,. N' Y gqg g f{'
Ibl.. . . ' a i.- h Y -e ,b- n b YOfh$ . pyh..;
CH5 CH10 CH1I CH4 CH12 )
( h j{ ' y .;;f: y lN .
h?
v g/_-m h.6W P5M .
,xe n,,-up specimen
&lgg t,
w m- mm m....,., AQh.
m .-
Alignment
.y S2 ypr m4[pp2T 4.2h. M [ Error h ]. .' Ii[g .
fl } -
m - ._
w C :
. CH14 CH15 CH13 CHI CH6 Figure 5-8 Charpy Impact Specimen Fracture Surfaces of the CPSES Unit No. 2 Reactor Vessel lleat-Affected-Zone (HAZ) Metal 5-25
Curve 764513 Temperature ( C) 0 50 100 150 200 250 300 110 i i i i i i i 100 -
700 Ultimate Tensile Strength 90 -
2- 600
$ 80 b
2 12 - . .
500 g 2
60 -
2 3-2 50 -
0.2% Yield Strength -
300 40 OA Unirradiated eA trradiated to a Fluence of 3.28 x 1018 n/cm2 80 Reduction in Area 2
70 -
n C i ~
60 -
C 50 -
.D 40 -
1 g Total Elongation O 30
, , 4
_ m -2 l 20 -
2 2 C C D 10 - -
4 Uniform Elongation l 1 1 I I
'. 0 100 200 300 400 500 600 l Temperature ( F) !
i 1
4 L
s Figure 5-9 Tensile Properties for CPSES Unit No. 2 Reactor Vessel Intennediate Shell Plate R3807-2 (IAngitudinal Orientation) 5 26
l Curve 764511 Temperature (*C) 0 50 100 150 200 250 300 110 , i i i , i i 100 -
Ultimate Tensile Strength - 700 90 3 2Q 600
.e A M2 -
$ 80 - u g E -
0 -
500$ 1 j 70
- v 2q 2 60 - v v _ 400 50 -
0.2% Yield Strength -
300 40 OA Unirradiated eA Irradiated to a Fluence of 3.28 x 1018 n/cm2 I 80 70 -
Reduction in Area
^
g " _
60 -
N l
[ 50 -
.D 40 -
j Total Elongation a 30 -
k
~
~
- 3 20 -
- R _ R 0 10
- ~ - O a Uniform Elongation g
i I I I I O 100 200 300 400 500 600
- Temperature ( F) 4 4
Figure 5-10 Tensile Propemes for CPSES Unit No. 2 Reactor VesselIntermediate Shell Plate R3807-2 (Transverse Orientation) 5-27 )
l e
Curve 784512 Temperature ( C) 0 50 100 150 200 250 300 110 i i i i i i Ultimate Tensile Strength - 700 100 -
90 -
1 i 2
80 -
.A g 2
500$
g 70 m
(_
60 -
"2 e 400 50 - 0.2% Yield Strength 40 oA Unirradiated eA Irradiated to a Fluence of 3.28 x 1018 n/cm2 80 70 -
2 _
C 60 f 2
~-
i 4
@ 50
= 40 -
g Total Elongation 3 30 -a_ u i u A. .
20 - 3 u2 i
0 -
c- a
" v 10 -
Uniform Elongation i I I I I O 100 200 300 400 500 600 Temperature ( F)
Figure 5-11 Tensile Properties for CPSES Unit No. 2 Reactor Vessel Weld Metal 5-28
4 I
i g ;, e -
, . c _. ,.3 . =..
x:.cacsa_
M 1
i i
k3 9^! 1 2 3 t '0 6 .76 4
[ .'.r.! d .!dddtit,!.r.,lMfj.p ddddddN .
J 3 1
1 1
,+
l' Specimen CL1 tested at 70F
, ,.. . ,. ~
- u: ~ee ... . , a;p,.,:r .-
i :: - ,~uxc3csR;,
i
%s9.} " i..2 ^3 M. T Y N 680 l 1
jddujdddd i .
j 4lg. ,
> :N-u
, : 5:. **; .y.i, ' , ,. ;' g
] @N yo . <_. , t :r. J1 a sc :
sd
... d
) .
i Specimen CL2 tested at 150F l
{
l
- 3r)h e.,;.
~
T -
j h 1 j Specimen CL3 tested at 550F i
i i
i
) Figure 5-12 Fractured Tensile Specimens from CPSES Unit No. 2 Reactor Vessel Intermediate j Shell Plate R3807 2 (longitudinal Orientation) j
,' 5-29 l
i
- ~a- .; & .s a;,yy y .;y
[j.-
- :0 % cacs y f'
1
$$.. ee' - 1 2.T44[ .Y crys,
] ,
[9. dddu f I M'-; _, .._ .
. g, Specimen CTI tested at 25cF Specimen CT2 tested at 15&F y Specimen CT3 tested at 55&F 1
Figure 5-13 Fractured Tensile Specimens from CPSES Unit No. 2 Reactor Vessel Intermediate Shell Plate R3807-2 (Transverse Orientation) 5-30
i e
l Specimen CW2 tested at 259 l
- r,r . . - ~~e,-~..
i.... . -. 3 . .
. :. c 305 R, .~ l
'8 e1l 12 s 4 4 . 's , ' .7._ s -
I i
i %%ichhHdddddsfdddf)di j i
i 4
g Specimen CW1 tested a: 1507
- g s -T i
- b'
~q i
t i
1 I
I i Specimen CW3 tested at 550P i
1 1
i f
i Figure 5-14 Fractured Tensile Specimens from CPSES Unit No. 2 Reactor Vessel Weld Metal 5-31
100.00 90.00-80.00-70.00-(1)
- 80.00-05 U
40.00-30.00-20.00- CL1 10.00- 70 F 0.00 . , , . .
0.00 0.10 0.20 0.30 STRAIN, IN/IN 100.00 90.00-80.00-70.00-u)
- 60.00-($
@ 50.00-0 g" 40.00- )
30.00-20.00- 2 10.00- 150 F 0.00 , , . . ,
0.00 0.10 0.20 0.30 STRAIN, IN/IN Figure 5-15 Engineering Stress-Strain Cmves for Intermediate Shell Plate R3807-2 Tensile l Specimens CLI and CL2 (longitudinal Orientation) 1 5-32 l l
1
1 i
i l
j l
I 1
I J
l 100.00 ;
I 90.00- . -
80.00- :
. 70.00-M -
Y. 80.00- .
50.00-H 40.00-M ,
i 30.00-20.00- CL3 10.00-
)
550 F 0.00 , , , , , l 0.00 0.10 0.20 0.30 i STRAIN, IN/IN l
l Figure 5-16 ' Engi-ring Stress-Strain Curve for Intermediate Shell Plate R3807-2 Tensile Specimen CIS (Imagitudinal O.ientation) 5-33
i.
l l
100.00 ;
i 90.00-80.00-70.00-to
- -80.00-I 50.00-
. ~
g 40.00-30.00- j-CT 1 20.00-25 F I 10.00_ ,
0.00 , , , '
0.00 0.10 0.20 O.30 STRAIN, IN/IN 100.00 90.00- :
P 80.00-
._ 70.00-co 60.00-50.00- f CC :
40.00-30.00-CT2 20.00_
10.00- 150 F -
0.00 , , ' '
0.00 0.10 O.NO O.30 STRAIN, IN/IN Figure 5-17 Engineering Stress-Strain Curves for Intermediate Shell Plate R3807-2 Tensile Specimens Crl and CT2 (Transverse' Orientation) i 5-34
100.00 90.00-80.00-1 70.00- .
W) 60.00- >
b g 50.00-(Z:
g 40.00-30.00- l CT3 20.00-10.00- 550 F l l 0.00 . . . . .
0.00 0.10 0.20 0.30 ,
STRAIN, IN/IN i 1
i 4
Figure 518 Engineering Stress-Strain Curve for Intermediate Shell Plate R3807-2 Tensile Specimen Cr3 (Transverse Orientation) 5-35 1
I
-100.00 ,
90.00-80.00-70.00-m
- 80.00- ,
50.00-40.00-30.00- -
20.00- CW2 10.00- 25 F 0.00 , . . . .
0.00 0.10 0.20 0.30 STRAIN, IN/IN 100.00 -
90.00- ,
80.00-
_ 70.00-u) i
- 60.00-Ui
@ 50.00-1 g 40.00-30.00-20.00-CW1 l
10.00- 150 F F
0.00 , , , , ,
0.00 0.10 0.20 0.30 I STRAIN, IN/IN t
Figure 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens CW2 and CW1 5-36
i
- l i
I 6
.100.00 ,
t N.00-80.00-70.00- '
(t) 60.00- ..
b y 50.00-12: '
40.00-
- j. 30.00- 3 20.00- -CW3 10.00- 550 F l' 1
O.00 , , , , ,
0.00 0.10 0.20 0.30 STRAIN, IN/IN
).
2 1
Figure 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen CW3 5 37
~
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY-6.1 Introduction Knowledge of the neutron environment within the mactor pressure vessel and surveillance capsule .
geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons.' First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test -
specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and futur: condition of the rsactor vessel, a relationship must be established between the neutron environment at various positions within the pressure vessel and that experienced b'y the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is generally derived solely from analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent
- years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between 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
- l. ' accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage
[ function for data correlation, ASTM Standard Practice E853. " Analysis and Interpretation of Light Water Reactor Surveillance Results." recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa j- function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing l- . Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials."
6-1
i i
}
This section provides the results of the neutron dosimetry evaluations performed in conjunction with' the analysis of test specimens contained in surveillance Capsule U, withdrawn at the end of the 1st
- fuel cycle. This evaluation is based on current state-of-the-art methodology and nuclear data. This -
repon provides a consistent up-to-date database for use in evaluating the material properties of the.
CPSES Unit 2 reactor vessel. l
-i I
' In each of the dosimetry evaluations, fast neutron exposure parameters in terms of neutron fluence j (E > l.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 t l
A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation l l
l capsules attached to the thermal shield are included in the reactor design to constitute the reactor !
t vessel surveillance program. The capsules are located at azimuthal angles of 58.5*,61.0*,121.5*,
238.5 ,241.0, and 301.5 relative to the core cardinal axis as shown in Figure 4-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1. The stainless i steel specimen containers are 1.182 by 1-inch anc' approximately 56 inches in height. The containers l are positioned axially such that the test specimens are centered on the core midplane, thus spanning p the central 5 feet of the 12 foot high reactor core. I In regard to the geometry depicted in Figure 4-1, it should be noted that, for the neutron pad arrangement in CPSES Unit 2, the azimuthal extent of the pad is not the same for all octants. For i octants containing no surveillance capsules the span of the neutron pad is 12.5"; while for octants !
containing surveillance capsule holders the span extends to 17.5"in order to position the capsules to i achieve the desired lead factors. Both of these pad configurations are considered in the fluence evaluations for the capsules and pressure vessel. f i
i From a neutronic standpoint, the surveillance capsules and associated suppon stmetures are significant. j The presence of these materials has a marked effect on both the spatial distribution of neutron flux and ,
the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel.
I 6 ;
r p
i E
h
i i
.i
. . . t
-In order to determine the neutron environment at the test specimen location, the capsules themselves j must be included in the analytical model.-
l 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), $(E > 0.1 MeV), and dpa/sec} through the vessel wall. The neutron f spectral information was required for the interpretation of neutron dosimetry withdrawn from the j surveillance capsules as well as for the determination of exposure parameter ratios; i.e., [dpa/sec]/[$(E l
> 1.0 MeV)], within the pressure vessel geometry. The relative radial gradient information was f
f
. 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. I The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, j
$(E > 1.0 MeV), at surveillance capsule positions and at several azimuthal locations on the pressure lt 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 j
^
cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the l i
locations of interest for each cycle of irradiation. They also established the means to perform similar i j :
l: predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core but also accounted for the effects of varying neutron yield per l
j fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel l
assemblies increased. -!
$ i j 'Ihe absolute cycle specific data from the adjoint evaluations together with the relative neutron energy j 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.
4- 3 -- Enable a direct comparison of analytical prediction with measurement.
i 6-3 l i
9 I
4- Establish a mechanism for projection of pressure vessel exposure as the design of each new -
fuel cycle evolves.
He forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R,0 geometry using the DORT two-dimensional discrete ordinates code Version 2.8.14"33 and V
the BUGLE 93 cross-section library . He BUGLE 93 library is a 47 energy group ENDF/B-VI
- based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was treated with a P3expansion of the scattering cross-sections and the angular discretization was modeled with an S, order of angular quadrature.
He core power distribution utilized in the reference forward transport calculation was derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inhereret in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Funhermore, 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 variations in peripheral power. Axial leakage effects were taken into account by application of an axial peaking factor of 1.2. 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; and, since an axial peak to average of 1.2 tends to be bound that observed in fuel assemblies with significant burnup, the use of this reference distribution is expected to yield somewhat conservative results. He degree of conservatism could range from approximately 10% for reactors employing an out/in fuel management strategy to as much as a factor of two for reactors that have transitioned to low-leakage fuel management.
All adjoint calculations were also carried out using an S, order of angular quadrature and the P3 cross-section approximation from the BUGLE 93 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 imponance functions for the exposure parameter of interest, in this case $(E > 1.0 MeV).
llaving the importance functions and appropriate core source distributions, the response of interest could be calculated as:
R(r,0) = f f f 1(r,0,E) S(r,0,E) r dr do dE r 0 E 6-4 i
. where: R(r,0) =; $(E > 1.0 MeV) at radius r and azimuthal angle 6.
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.
i Although the adjoint imponance functions used in this analysis were based on a response function ,
{
defined by the threshold neutron flux $(E > 1.0 McV), prior calculations"* have shown that, while the' !
I implementation of low leakage loading patterns significantly impacts both the magnitude and spatial j distribution of the neutron field, changes in the relative neutron energy spectrum are of second order.
l Thus, for a given location the ratio of [dpa/sec]/[$(E > 1.0 MeV)] is insensitive to changing core i source distributions. In the application of these adjoint importance functions to the CPSES Unit 2 l 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 $(E > !
1.0 MeV) solutions from the individual adjoint evaluations. {
I i
~
The reactor core power distributions, including the cycle specific axial distribution, used in the plant specific adjoint calculations were taken from the fuel cycle design report for the first operating cycle j of CPSES Unit 2im, 1 Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the.
~ capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the pressure vessel wall.
In Table 6-1, the calculated exposure parameters [$(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec] are l given at the geometric center of the two surveillance capsule positions for both the reference and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, j
- are meant to establish the absolute comparison of measurement with analysis. The reference data derived from the forward calculation are provided as a conservative exposure evaluation against which plant specific fluence calculations can be compared. Similar data are given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for the reference and cycle one plant specific power distributions. It is important to note that the data for the vessel 6-5 j
inner radius were taken at the clad / base metal interface; and, thus, represent the maximum predicted exposure levels of the vessel wall itself.-
i l
Radial gradient information applicable to $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec is given in Tables 6-3,6-4, and 6-5, respectively. The data, obtained from the reference forward neutron transport l calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Tables 6-3 i through 6-5.
For example, the neutron flux $(E > 1.0 MeV) at the 1/4T depth in the pressure vessel wall along the 45' azimuth is given by:
$,,g45*) = 4(220.27, 45 ) F(225.75, 45*)
where: $yy(45*) = Projected neutron flux at the 1/4T position on the 45 azimuth.
$(220.27,45 ) = Projected or calculated neutron flux at the vessel inner radius on the 45 azimuth.
F(225.75,45') = Ratio of the neutron flux at the 1/4T position to the flux at the vessel inner radius for the 45 azimuth. This data is obtained from Table 6-3 Similar expressions apply for exposure parameters expressed in terms of $(E > 0.1 MeV) and dpa/sec where the attenuation function F is obtained from Tables 6-4 and 6-5, respectively.
As noted earlier in this Section, the neutron pad arrangement in CPSES Unit 2 is not the same for all octants of the reactor. For the analysis of the neutron flux to the pressure vessel, the DORT calculations were performed with the neutron pad extending from 32.5'- 45.0 . This configuration produces the maximum flux to the pressure vessel.
6.3 Neutron Dosimetry The passive neutron sensors included in the CPSES Unit 2 surveillance program are listed in Table 6-
- 6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that 6-6
were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters of interest [$(E > 1.0 MeV),
$(E > 0.1 MeV), dpa/sec]. The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium shielded uranium and neptunium fission monitors were accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target t
material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:
l 1
- The measured specific activity of each monitor.
The physical characteristics of each monitor.
l The operating history of the reactor.
- The energy response of each monitor.
The neutron energy spectrum at the monitor location.
The specific activity of each of the neutron monitors was determined using established ASTM procedures"' *""# 3" Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectremeter. The irradiation history of the CPSES Unit 2 reactor during cycle one was supplied by NUREG-0020 " Licensed Operating Reactors Status Summary Report," for the applicable period. The irradiation history applicable to capsule U is given in Table 6-7.
Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full power operation were determined from the following equation:
A 1 R= !
P . x, l
'][e . x,']
N, F Y [ p C, [1 -e ret 1
6-7
where:
R = . Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,,, (rps/ nucleus).
A = Measured specific activity (dps/gm).
No = Number of target element atoms per gram of sensor.
F = Weight fraction of the target isotope in the sensor material.
Y = Number of product atoms produced per reaction.
P, = Average core power level during irradiation period j (MW). j P, r = Maximum or reference power level of the reactor (MW). .
C3 = Calculated ratio of $(E > 1.0 MeV) during irradiation period j to the time weighted average $(E > 1.0 MeV) over the entire irradiation period.
A = Decay constant of the product isotope (1/sec).
t, = Length of irradiation period j (sec).
t, = Decay time following irradiation period j (sec).
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
In the equation describing the reaction rate calculation, the ratio [P;]/[P,,,) accounts for month by month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio C,, which can be calculated for each fuel cycle using the adjoint transport technology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single cycle irradiation C3 is normally taken to be 1.0. However, for multiple cycle irradiations, panicularly those employing low leakage fuel management, the additional C; term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.
For the irradiation history of Capsule U, the flux level term in the reaction rate calculations was developed from the plant specific analysis provided in Table 6-1. Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Table 6-8.
6-8
i Values of key fast neutron exposure parameters were derived from the measured" reaction rates using the FERRET least squares adjustment coder 32i, The FERRET approach used the measured reaction rate
. data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust i the group fluxes from the tnal spectrum to produce a best fit (in a least squares sense) to'the measured - l reaction rate data. 'Ihe " measured" exposure parameters along with the associated uncertainties were -
l then obtained from the adjusted spectrum.
In the FERRET evaluations, a log-normal least squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the ,
measured values f are linearly related to the flux $ by some response matrix A:
1
~
f = A $"' _
l 1
i where i indexes the measured values belonging to a single data set s, g designates the energy group,
and a delineates spectra that may be simultaneously adjusted. For example, )
l R=[ois i , $s relates a set of measured reaction rates R, to a single spectrum $, by the multigroup reaction cross-section oi,.. The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were approximated in a multi group format consisting of 53 energy groups. The trial input spectrum was converted to the FERRET 53 group structure using the SAND-Il code"31 This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET.
The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry file"51, were also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectmm, as expanded to 620 groups, was employed as a weighting function in the cross-section collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix 6-9 l
l l
for each sensor reaction were also constructed from the information contained on the ENDF/B-VI data files. These matrices included energy group to energy group uncenainty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were not included. The omission of this additional uncertainty information does not significantly impact the results of the adjustment.
Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation was taken from the center of the surveillance capsule modeled in the reference forward transport calculation. While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the ENDF/B-VI data files, the covariance matrix for the input trial spectrum was constructed from the following relation:
Mi=R +RRiPi se n a s at where R,, specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set of values. Tne fractional uncertainties R, specify additional random uncertainties for group g that are conelated with a correlation matrix given by:
P , = [1-0] 5 , + 0 e ~"
88 88 where: ,
H= 2 2y The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (6 specifies the strength of the latter term). The value of 5 is I when g = g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerker'30 Maerker's results j are closely duplicated when y = 6.
The uncertainties associated with the measured reaction rates included both statistical (counting) and ;
systematic components. The systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, inadiation history corrections, and corrections for competing reactions 6-10 I i
i in the individual sensors. A combination of these sources of uncertainty result in reaction rate .i 1
uncertainties of15% for non-fission reactions and 10% for the fission reactions.
.]
l Results of the FERRET evaluations of the Capsule U dosimetry are given in Table 6-9. He data j
. summarized in this table includes fast neutron exposure evaluations in terms of 4(E > 1.0 MeV),4(E . I
> 0.1 MeV), and dpa. In general good results were achieved in the fits of the adjusted spectra to the individual measured reaction rates. The measured and FERRET adjusted reaction rates for each.
reaction are given in Table 6-10. Table 6-10 also depicts the level of precision associated with each j reaction via a ratio of calculated rate to measured rate. De adjusted spectra from the least squares l evaluation is given in Table 6-11 in the FERRET 53 energy group suucture. Table 6-12, titled ;
" Comparison of Calculated and Measured Neutron Exposure Levels for CPSES Unit 2 Surveillance Capsule U", compares the measured and calculated fluence at the capsule. The results for capsule U l
are consistent with results obtained from similar evaluations of dosimetry from other Westinghouse reactors. ;
I 6.4 Proiections of Pressure Vessel Exposure
.i, Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6- !
i
- 13. Along with the current (0.904 EFPY) exposure, projections are also provided for exposure periods l
- of 15 EFPY,32 EFPY and 48 EFPY. In computing these vessel exposures, the calculated values from Table 6-13 were scaled by the average measurement / calculation ratios (M/C) observed from the evaluations of dosimetry from Capsule U for each fast neutron exposure parameter. This procedure resulted in bias factors of 1.091,1.225, and 1.148 being applied to the calculated values of @(E > 1.0 (
MeV), @(E > 0.1 MeV), and dpa, respectively. Projections for future operation were based on the l t
assumption that the best estimate exposure rates characteristic of the first cycle irradiation would [
continue to be applicable throughout plant life. !
i The overall uncertainty in the best estimate exposure projections within the pressure vessel wall stem ,
t primarily from two sources:
- 1) the uncertainty in the M/C ratios derived from the plant specific measurement data base; l r
and l
- 2) the analytical uncertainty associated with relating the results at the measurement locations l to the desired results within the pressure vessel wall.
l l
6-11 h
l
, Uncertainty in the M/C ratios derives directly from the individual uncenainties in the measurement process, in the least squares adjustment procedure, and in the location of the surveillance capsule and-j cavity dosimetry sensor sets. The analytical uncertainty in the relationship between the exposure of the pressure vessel and the exposure at the measurement locations is and on downcomer water density variations and vessel inner radius tolerance relative to the surveillance capsule data.
The 1o uncenainties associated with the M/C ratios applicable to @(E > 1.0 MeV),
d .
j @(E > 0.1 MeV), and dpa are given in Table 6-9 of this report. The additional information peninent
- . to the required analytical uncenainty for vessel locations has been obtained from benchmarking studies using the Westinghouse neutron transpon methodology and from several comparisons of power reactor g
- mtemal surveillance capsule dosimetry and reactor cavity dosimetry for which the irradiation history of
- ' all sensors was the same.
j Based on these benchmarking evaluations the additional uncertainty associated with the tolerances in 3
dosimetry positioning, vessel thickness, vessel inner radius and downcomer temperature was estimated to be approximately 6% for all exposure parameters. These uncenainty components were then j combined as follows:
f Io UNCERTAINTY ;
i
@(E > 1.0 MeV) @(E > 0.1 MeV) dga i M/C Ratio 8% 16% 11 %
Analytical 6% 6% 6%
Combined 10.0 % 17 % 13 %
. Thus, the total uncertainty associated with the neutron exposure projections at the pressure vessel clad / base metal interface for CPSES Unit 2 was estimated to be:
10 Uncenainty
@(E > 1.0 MeV) 10%
@(E > 0.1 MeV) 17 %
1 1
dpa 13 %
These uncenainty values are well within the 20% 1o uncenainty in vessel fluence projections required by the FTS rule.
6-12 )
_ _ __a
In the calculation of exposure gradients for the CPSES Unit 2 reactor vessel, exr50sure projections to 15,32, and 48 EFPY were also employed. 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 6-14.
In order to access RTm vs fluence curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations:
dpa(1/47)
$(1/47) = $(07) -
dpa(07) and dra(3/47)
$(3/47) = $(07) dpa(07)
Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Table 6-15 updated lead factors are listed for each of the CPSES Unit 2 surveillance capsules. Lead factor data l based on the accumulated fluence for cycle one are provided for each remaining capsule.
I l
6-13
. FIGURE 6-1 PLAN VIEW OF A DUAL REACTOR VESSEL SURVEILLANCE CAPSULE l
i l
i
- 58.5= - 61.0*
! Fe Cu
. .\
,O - 81.625 in.
., r 5 \
l NEUTRON PAD 6-14
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER
$ (E >l.0 MeV) [n/cm2-sec]
LOCATION 21Q* 1LJ' CYCLE 1 9.64E+10 1.06E+11 CRSD Data 1.20E+11 1.28E+11
$ (E > 0.1 MeV) [n/cm2-sec)
LOCATION 21Q* 1Li' CYCLE 1 4.05E+11 4.49E+11 CRSD Data 5.02E+11 5.43E+11 fron Displacement Rate [dpa/sec)
LOCATION 210* 11.J' CYCLEI 1.84E-10 2.02E-10 CRSD Data 2.28E-10 2.44E-10 l
6-15
TABLE 6-2 CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE .
2
$(E > 1.0McV) [n/cm -sec]
Of* 15.0* 21,q' 30 0* 310' 45.0*
CYCLE 1 1.26E+10 1.93E+10 2.33E+10 2.31E+10 2.12E+10 2.58E+10 CRSD Data 1.61E+10 2.51E+10 2.97E+10 2.86E+10 2.53E+10 2.99E+10 2
&(E > 0.lMeV) [n/cm -sec] !
0 O,0* 15.0* 25.0* 30.0* 35.0* 45.0*
' CYCLE 1 2.53E+10 3.93E+10 4.80E+10 4.81E+10 4.92E+10 6.15E+10 CRSD Data 3.23E+10 5.llE+10 6.11E+10 5.94E+10 5.88E+10 7.15E+10 Iron Atom Displacement Rate [dpa/sec)
D.Q* 15.0* 25 0' 30 0* 35.0* 45.0*
CYCL.EI 1.94E-11 2.95E-11 3.55E-11 3.53E-11 3.30E-11 4.02E-11 CRSD Data 2.48E-11 3.83E-11 4.52E-11 4.36E-11 3.95E 11 4.68E-11 4
6-16 ;
l l
l
TABLE 6-3 )
RELATIVE RADIAL DISTRIBUTION OF $(E > 1.0 MeV) ,
l WITHIN THE PRESSURE VESSEL WALL i
Radius (p.g1L 0. O' M M M 45.0*
220.35* 1.00 1.00 1.00 1.00 1.00 221.00 0.959 0.958 0.958 0.954 0.957 222.30 0.852 0.850 0.850 0.841 0.846 223.60 0.739 0.736 0.738 0.728 0.730 224.89 0.634 0.630 0.630 0.623 0.623 225.87 0.562 0.557 0.556 0.551 0.548 227.01 0.487 0.482 0.481 0.474 0.473 228.63 0.395 0390 0.390 - 0.385 0.381 230.90 0.326 0.321 0.320 0.318 0.312 231.39 0.274 0.269 0.268 0.266 0.260 232.68 0.229 0.225 0.224 0.222 0.217 234.14 0.188 0.184 0.183 0.181 0.176 l 235.56 0.150 0.146 0.146 0.144 0.140 236.90 0.127 0.124 0.124 0.124 0.118 l
237.88 0.110 0.107 0.107 0.106 0.102 ;
239.18 0.091 0.089 0.088 0.087 0.083 l
J 240.47 0.075 0.073 0.072 0.072 0.068 241.77 0.061 0.058 0.058 0.058 0.054 242.42* 0.058 0.055 0.055 0.055 0.05i NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius i
l I
6-17 J
. TABLE 6 RELATIVE RADIAL DISTRIBUTION OF $(E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius
(.CD1)_ 0. O' 15 0' M ),10*
Q 45.0' 220.35"' 1.00 1.00 1.00 1.00 1.00 221.00 1.00 1.00 1.00- 1.00 1.00 222.30 - 1.00 0.997 0.997 0.990 0.989 223.60 0.967 0.958 0.958 0.951 0.945 -
224.89 0.922 0.909 0.909 0.902 0.893 225.87 0.884 0.869 0.869 0.863 0.850 227.01 0.837 0.821 0.821 0.814 0.800 228.63 0.770 0.753 0.753 0.748 0.729 230.09 0.710 0.692 0.693 0.689 0.667 231.39 0.657 0.639 0.639 0.635 0.612 232.68 0.604 0.587 0.586 0.583 0.558 234.14 0.546 0.529 0.529 0.526 0.500 235.76 0.484 0.469 0.468 0.467 0.438 236.90 0.441 0.427 0.427 0.426 0.396 237.88 0.403' O.391 0.390 0.389 0.360 239.18 0.356 0.345 0.344 0.343 0.313 240.47 0.310 0.300 0.298 0.298 0.267 241.77 , 0.265 0.252 0.251 0.251 0.220 242.42(23 0.256 0.242 0.240 0.240 0.210 l
1 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-18 l
. TABLE 6-5 RELATIVE RADIAL DISTRIBUTION OF dpa/sec WITHIN THE PRESSURE VESSEL WALL-Radius (EEL 0. O' 15.0* 21,Q* 35 0* 45 0*
220.35"' l.00 1.00 1.00 1.00 1.00 221.00 0.965 0.964 0.965 0.963 0.965 222.30 0.877 0.876 0.877 0.873 0.878 223.60 0.785 0.782 0.784 0.784 0.788 224.89 0.699 0.695 0.697 0.699 0.703 225.87 .0.638 0.634 0.636 0.641 Of42 227.01 0.575 0.571 0.573 0.578 0.579 228.63 0.495 0.491 0.493 0.499 0.500 230.09 0.432 0.427 0.430 0.438 0.437 231.39 0.383 0.378 0.380 0.389 0.387 232.68 0.339 0.334 0.336 0.345 0.342 234.14 0.295 0.291 0.292 0.300 0.297
-235.76 0.252 0.248 0.250 0.258 0.253 236.90 0.224 0.221 0.222 0.231 0.224 237.88 0.202 0.199 0.200 0.208 0.201 239.18 0.175 0.173 0.174 0.180 0.172 240.47 0.150 0.148 0.149 0.I '5 0.145 241,77 0.128 0.124 0.124 0.130 0.120 242.42 ' O.124 0.119 0.119 0.125 0.115 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-19
TABLE 6-6 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS i
Reaction Target Fission Monitor of Weight Response Product Yield Matenal Interest Fraction Ranoe - Half-Life (%)
Copper Cu(n.a )Co" 0.6917 E > 4.7 MeV 5.271 yrs Iron Fe"(n.p)Mn" 0.0580 E > 1.0 MeV 312.5 days Nickel Ni"(n,p)Co" 0.6827 E > 1.0 MeV - 70.78 days
! - Uranium-238* U2"(n,f)Cs'" 1.0 E > 0.4 MeV 30.17 yrs 6.00 Neptunium-237*' Np2 "(n f)Cs'" 1.0 E > 0.08 MeV 30.17 yrs 6.27 Cobalt-Aluminu'n* Co"(n,y)Co* 0.0015 0.4ev>E>0.015 MeV 5.271 yrs
. , Cobalt-Aluminum Co"(n,y)Co" 0.0015 E > 0.015 MeV 5.271 yrs L
i
- Notes
- 1)
- Denotes that sensor is cadmium shicided.
- 2) Nuclear data for sensor materials were taken from the following ASTM Standards:
ASTM-E1005-84 (91)
ASTM-E7N-90 ASTM-E705-90 6-20
- l. t l
TABLE 6-7 :
4 MONTHLY THERMAL GENERATION DURING THE FIRST FUEL CYCLE OF THE CPSES UNIT 2 REACTOR !
t Thennal Generation Year Month (MW-hr) l
! 1993 Mar 13,098 ,
Apr 665,234 ;
May 499,370 Jun 0 Jul 1,398,238 Aug 2,327,446 l Sep 1,317,156 Oct 2,397,394 !
Nov 2,063,460 Dec 2,512,502 1994 Jan 2,471,383 !
Feb 1,545,139 t Mar 1,165,937 ;
Apr 1,225,908 ;
May 0 t Jun 608,844 '
Jul 2,417,832 Aug 1,728,622 l Sep 2,323,594
- Oct 342,494 l i
i
)
i
)
l
. 1 6-21
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE U SATURATED ACTIVITIES AND REACTION RATES Measured Saturated Reaction Activity Activity Rate Reaction (DPS/GM) (DPS/GM) (DPS/ ATOM)
Cu-63 (n,a) Co-60 95-218 Top 4.78E+N 4.59E+05 7.01E-17 95 223 Mid 4.28E+04 4.11E+05 6.27E-17 95-228 Bottom 4.29E+04 4.12E+05 6.29E-17 '
" AVERAGES " 4.45E+04 4.27E+05 6.52E-17 Fe-54 (n.p) Mn-54 95-220 Top 1.39E+06 4.11E+06 6.57E-15 95-225 Mid 1.27E+06 3.75E+06 6.00E-15 95-230 Bottom 1.27E+06 3.75E+06 6.00E-15
- AVERAGES *
- 1.31E+06 3.87E+06 6.19E-15 Ni-58 (n.p) Co-58 95-219 Top 1.30E+07 6.07E+07 8.67E-15 95-224 Middle 1.21E+07 5.65E+07 8.07E-15 95-229 Bottom 1.21E+07 5.65E+07 8.07E-15
- AVERAGES " 1.24E+07 5.79E+07 8.27E-15 Co-59 (n.y) Co-60 95-216 Top 9.83E+06 9.44E+07 6.16E-12 95-221 Middle 1.03E+07 9.89E+07 6.46E-12 95-226 Bottom 1.03E+07 9.89E+07 6.46E-12
" AVERAGES ** 1.01E+07 9.74E+07 6.36E-12 6-22
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE U SATURATED ACTIVITIES AND REACTION RATES -
Measured Saturated Reaction Activity Activity Rate Reaction (DPS/GM) (DPS/GM) (DPS/ ATOM)
Co-59 (n,y) Co-60*
95-217 Top 5.29E+06 5.08E+07 - 3.32E-12 95-222 Middle 5.65E+06 5.43E+07 3.54E-12 95-227 Bottom 5.55E+06 5.33E+07 3.48E-12
- AVERAGES " 5.50E+06 5.28E+07 3.45E-12 U-238 (n,0 Cs-137*
95-214 Middle 1,37E+05 6.75E+06 3.88 E-14 *
- Np-237 (n,0 Cs-137*
95-215 Middle 1.21E+06 5.97E+07 3.75E-13 -
- Denotes that monitor is cadmium shielded.
" Corrected for Pu-239 build-in at .87 l i
i h
)
I 1
l 6-23
TABLE 6-9
SUMMARY
OF NEUTRON DOSIMETRY RESULTS SURVEILLANCE CAPSULE U EgL Fluence Ouantity 2 (n/cm ,3,c) gigggj2 } Uncertainty FLUX. E< 0.414 EV 1.19E+11 3.41E+18 122 %
FLUX E> 0.1 MEV 5.50E+11 1.57E+19 216 %
FLUX. E> 1.0 MEV 1.15E+11 3.28E+18 28%
DPNSECOND 2.32E-10 6.61E-03 ' 11 %
D 6-24 i
?
?
TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED ,
REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULE U 1 $
REACTION RATE RATIO (DPS/ NUCLEUS) CALC / MEAS Reaction Mgm Adi Cale Adi Cale l
- Cu63(n a) Co60 6.52E-17 6.35E-17 0.97 Np237(n.f) Csl37' 3.74E-13 3.66E-13 0.98 Fe54(n.p) Mn54 6.19E-15 633E-15 1.02 l NiS8(n.p) CoS8 8.27E-15 8.57E-15 1.04 U238(n,f) Cs137' 3.88E-14 3.50E-14 0.90 !
CoS9(n,y) Co60 6.36E-12 6.28E-12 0.99 CoS9(n,y) Co60* 3.44E-12 3.47E-12 1.01 ,
i
- Denotes that monitor is cadmium shielded. ;
8 r
l 6-25
r .
l l
l' TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE U Group # Enerev (MeV) Flux (n/cm 2-sec) Group # Energy (MeV) Flux (n/cm 2-sec)
I 1.73E+01 8.63E+06 28 9.12E-03 2.48E+10 2~ 1.49E+01 1.83E+07 29 5.53E-03 3.19E+10 3 1.35E+01 6.64E+07 30 336E-03 9.93E+09 4 1.16E+01 1.79E+08 31 2.84E-03 9.51E+09 5 1.00E+01 3.99E+08 32 2.40E-03 9.24E+09 6 8.61E+00 6.85E+08 33 2.04E-03 2.71E+10 7 7.41E+00 1.64E+09 34 1.23E-03 2.63E+10 8 6.07E+00 2.46E+09 35 7.49E-04 2.42E+10 9 4.97E+00 5.05E+09 36 4.54E-04 2.18E+10 10 3.68E+00 5.92E+09 37 2.75E-04 2.40E+10 11 2.87E+00 1.08E+10 38 1.67E-04 2.50E+10 .
12 2.23E+00 1.73E+10 39 1.01E-04 2.53E+10 13 1.74E+00 2.45E+10 40 6.14E-05 2.51E+10 14 1.35E+00 2.87E+10 41 3.73E-05 2.47E+10 15 1.l lE-00 5.10E+10 42 2.26E-05 2.40E+10 16 8.21E-01 6.05E+10 43 1.37E-05 2.32E+10 17 6.39E-01 6.68E+10 44 8.32E-06 2.24E+10 18 4.98E-01 4.60E+ 10 45 5.04 E-06 2.15E+10 19 3.88E-Ol 6.99E+10 46 3.06E-06 2.13E+10 20 3.02E-01 7 16E+10 47 1.86E-06 2.12E+10 21 1.83E-01 7.29E+10 48 1.13E-M 1.59E+10
.22 1.11E-01 5.33E+10 49 6.83E-07 2.00E+10 23 6.74E-02 4.14E+10 50 4.14E-07 2.20E+10 24 4.09E-02 2.23E+10 51 2.51E-07 2.12E+10 25 2.55E-02 2.58E+10 52 1.52E-07 1.89E+10 26 1.99E-02 1.24E+10 53 9.24E-08 5.74E+10 27 1.50E-02 2.16E+10 Note: Tabulated energy levels represent the upper energy in each group.
h 6-26 I i
. _ . _ ~ _ . - _ _ . _ . . _ . _ . _ _ _ _.
TABLE 6-12 !
COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE ,
LEVELS FOR CPSES UNIT 2 SURVEILLANCE CAPSULE U j l
Calculated Measured M/_C l Fluence (E > 1.0 MeV) [n/cm'-sec) 3.01E+18 3.28E+18 1.09 j Fluence (E > 0.1 MeV) [n/cm'-sec) 1.28E+19 1.57E+19 1.23 l dpa 5.76E-03 6.61E-03 1.15 l
l 1
1 l
l l
l 6-27 1
4
TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON TIIE PRESSURE VESSEL CLAD / BASE METAL INTERFACE BEST ESTIMATE EXPOSURE (0.904 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E 15.0" 25.0" 30.0* 5 2.5' 41' E > 1.0 3.92E+17 6.02E+17 7.26E+17 7.20E+17 6.59E+17 8.02E+17 E > 0.1 8.84E+17 1.38E+18 1.68E+18 1.68E+18 1.72E+18 2.15E+18 dpa 6.35E44 9.66E-04 1.16E-03 1.16E-03 1.08E-03 1.32E-03 BEST ESTIMATE EXPOSURE (15 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E 15.0* 21Q* 30.0* E 4.f E > 1.0 6.51E+18 9.98E+18 1.20E+19 1.19E+19 1.09E+19 1.33E+19 E > 0.1 1.47E+19 2.28E+19 2.78E+19 2.79E+19 2.85E+19 3.56E+19 dpa 1.05E-02 1.60E-02 1.93E-02 1.92E-02 1.79E-02 2.19E-02 BEST ESTIMATE EXPOSURE (32 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E 110 210 3E 31' E E > 1.0 1.39E+19 2.13E+ 19 2.57E+19 2.55E+19 2.33E+19 2.84E+19 E > 0.1 3.13E+19 4.87E+19 5.93E+19 5.95E+19 6.08E+19 7.61E+19 dpa 2.25E-02 3.42E-02 4.12E-02 4.09E-02 3.83E-02 4.66E-02 BEST ESTIMATE EXPOSURE (48 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E 15.0* 21Q* Q 2Q_0* 3ia 45*
E > 1.0 2.08E+19 3.19E+19 3.86E+19 3.82E+19 3.50E+19 4.26E+19 E > 0.1 4.69E+ 19 7.30E+19 8.90E+19 8.92E+19 9.12E+19 1.14E+20 dpa- 3.37E-02 5.13E-02 6.I7E-02 6.14E-02 5.74E-02 6.99E-02 6-28
TABLE 6-14 NEUTRON EXPOSURE VALUES FOR THE CPSES UNIT 2 REACTOR VESSEL l
FLUENCE BASED ON E > 1.0 MeV SLOPE ODEG 15 DEG 25 DEG 30 DEG 35 DEG 45 DEG 15 EFPY FLUENCE SURFACE 6.51E+18 9.9811+ 18 1.20E+19 1.19E+19 1.09E+19 1.33E+19 1/4T 3.66E+18 5.56E+18 6.70E+18 6.62E+18 6.03E+18 7.29E+18 3/4T 8.26E+17 1.24E+18 1.49E+18 1.47E+18 1.36E+18 1.57E+18 32 EFPY FLUENCE SURFACE 1.39E+19 2.13E+19 2.57E+19 2.55E+19 2.33E+19 2.84E+19 1/4T 7.80E+18 1.19E+19 1.43E+19 1.41E+19 1.29E+19 1.55E+19 3/4T 1.76E+18 2.64E+18 3.19E+18 3.13E+18 2.89E+18 3.35E+18 48 EFPY FLUENCE SURFACE 2.08E+19 3.19E+19 3.86E+19 3.82E+19 3.50E+19 4.26E+19 1/4T 1.17E+19 1.78E+19 2.14E+19 2.12E+19 1.93E+19 2.33E+19 3/4T 2.64E+18 3.96E+18 4.78E+18 4.70E+18 4.34E+18 5.02E+18 FLUENCE BASED ON dpa SLOPE ;
i 0DEG 15 DEG 25 DEG 30 DEG 35 DEG 45 DEG 1
15 EFPY FLUENCE SURFACE 6.51E+18 9.98E+18 1.20E+19 1.19E+19 1.09E+19 1.33E+19 1/4T 4.15E+18 6.33E+18 7.66E+18 7.59E+18 7.01E+18 8.54E+18 3/4T 1.46E+18 2.21E+18 2.67E+18 2.66E+18 2.53E+18 2.67E+18 32 EFPY FLUENCE SURFACE 1.39E+19 2.13E+19 2.57E+19 2.55E+19 2.33E+19 2.84E+19 1/4T 8.86E+18 1.35E+19 1.63E+19 1.62E+19 1.50E+19 1.82E+19 3/4T 3.1IE+18 4.71E+18 5.71E+18 5.68E+18 5.39E+18 5.70E+18 48 EFPY FLUENCE SURFACE 2.08E+19 3.19E+19 3.86E+19 3.82E+19 3.50E+19 4.26E+19 1/4T 1.33E+19 2.03E+19 2.45E+19 2.43E+19 2.24E+19 2.73E+19 3/4T 4.66E+18 7.06E+18 8.56E+18 8.52E+18 8.09E+18 8.55E+18 i
l 6-29
l
.,- l TA.BLE 6-15 l:
UPDATED LEAD FACTORS FOR CPSES UNIT 2 ;
SURVEILLANCE CAPSULES i CAPSULE LEAD FACTOR l V. 3.74 i U 4.10* ;
i X 4.10
?
W 4.10 ' ,
s .>
- 4. Y 3.74 !
Z 4.10 i l
- WITHDRAWN EOC 1 !
i
~!
i i
f
?
r e
.6-30
l l
SECTION 7.0 :
RECOMMENDED SURVEILLANCE CAPSULE REMOVAL SCHEDULE !
The following surveillance capsule removal schedule meets the requirements of ASTM E185-82 and is i recommended for future capsules to be removed from the CPSES Unit No. 2 reactor vessel ;
i TABLE 7-1 i Recommended Survetilance Capsule Removal Schedule for the CPSES Unit No. 2 >
- Reactor Vessel Capsule location Withdrawal Fluence Capsule (degree) Lead Factor EFPYd (n/cm', E > 1.0 MeV)
If" 58.5 4.098 0.904 3.28 x 10 ,
X 238.5 4.098 7.805 2.836 x 10 i W 121.5 4.098 11.701 4.255 x 10 "4 2
- Z 301.5 4.098 Standby ----- 4 V 61.0 3.744 Standby Y 241.0 3.744 Standby --- -
$U IY.b ,
(a) Effective Full Power Years (EFPY) from plant startup.
(b) Actual measured neutron fluence (c) Approximately equal to the projected vessel fluence at 48 EFPY.
V i'
l 4
4 7-1
,- .. . . . _ . _ . . . = -
(
~ SEC'I1ON
8.0 REFERENCES
- 1. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S.
Nuclear Regulatory Commission, May,1988. l l
- 2. Code of Federal Regulations,10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", . i
. U.S. Nuclear Regulatory Commission, Washington, D.C. .
- 3. . WCAP-10684, " Texas Utilities Generating Company Comanche Peak Unit No. 2 Reactor Vessel !
Radiation Surveillance Program", L. R. Singer, October 1984.
I
- 4. Section 111 of the ASME Boiler and Pressure Vessel Code, Appendix G, " Protection Against r Nonductile Failure". !
- 5. ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Determine ,
Nil-Ductility Transition Temperature of Ferritic Steels",in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA. ,
- 6. Code of Federal Regulations,10 CFR Part 50, Appendix H. " Reactor Vessel Material j Surveillance Program Requirements". U.S. Nuclear Regulatory Commission, Washington, D.C. j
- 7. ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled
- Nuclear Power Reactor Vessels", E706 (IF), in ASTM Standards, Section 3, American Society for ;
Testing and Materials, Philadelphia, PA,1993.
- 8. ASTM E23-93a, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials",
in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.
l
- 9. ASTM A370-92, " Standard Test Methods and Definitions for Mechanical Testing of Steel j 3
Products", in ASTM Standards, Section 3 American Society for Testing and Materials, j Philadelphia, PA,1993. ;
- 10. ASTM E8-93, " Standard Test Methods for Tension Testing of Metallic Materials", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
11, ASTM E21-92, " Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
- 12. ASTM E83 93, " Standard Practice for Verification and Classification of Extensometers",in ;
ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
81
- 13. RSIC Computer Code Collection CCC-543, " TORT-DORT Two- and Three-Dimensional Discrete Ordinates Transpon" Version 2.8.14", January 1994.
- 14. RSIC Data Library Collection DI.C-175, " BUGLE-93, Production and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron / Photon Cross-Section Libraries Derived from ENDF/B VI Nuclear Data", April 1994.
- 15. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.
- 16. Nuclear Science and Engineering, Volume 94, " Accounting for Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis", R. E. Maerker, et al, Pages 291-308, 1986.
- 17. WCAP 10804-RI, "The Nuclear Design and Core Physics Characteristics of the Comanche Peak Unit 2 Nuclear Power Plant Cycle 1", P. J. Sipush, March 1993.
- 18. ASTM Designation E482-89, " Standard Guide for Application of Neutron Transpon Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 19. ASTM Designation E560-84, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 20. ASTM Designation E693-79," Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 21. ASTM Designation E706-87, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards",in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 22. ASTM Designation E853-87, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993,
- 23. ASTM Designation E261-90, " Standard Practice for Determining Neutron Fluence Rate, Fluence, and Spectra by Radioactivation Techniques", in AS'IM Standards, Section 12. American Society for Testing and Materials, Philadelphia, PA,1993.
8-2
- 24. ASTM Designation E262-86, " Standard Method for Determining Thermal Neutron Reaction and j Fluence Rates by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials Philadelphia, PA,1993. l i
- 25. ASTM Designation E263-88, " Standard Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and j
~
Materials, Philadelphia, PA,1993. I 1
- 26. ASTM Designation E264-92, " Standard Method for Measuring Fast-Neutron Reaction Rates by l l Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and
- j. Materials, Philadelphia, PA,1993.
i
- 27. ASTM Designation E481-92, " Standard Method for Measuring Neutron-Fluence Rate by. ,
i Radioactivation of Cobalt and Silver",in ASTM Standards, Section 12, American Society for
?
Testing and Materials, Philadelphia, PA,1993.
i 28. ASTM Designation E523-87, " Standard Test Method for Measuring Fast-Neutron Reaction Rates 1 by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
. 29. ASTM Designation E704-90, " Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing i and Materials, Philadelphia, PA,1993.
i I
- I
! 30. ASTM Designation E705-90, " Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 31. ASTM Designation E1005 84, " Standard Test Method for Application and Analysis of ;
- Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, l American Society for Testing and Materials, Philadelphia, PA,1993.
- 32. IIEDieTME 79-40, " FERRET Data Analysis Core", F. A. Schmittroth Hanford Engineering j Development Laboratory, Richland, WA, September 1979.
4 N
- 33. AFWL "IR-7-41. Vol. I-IV, "A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation", W. N. McElroy, S. Berg and T. Crocket, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1%7.
- 34. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Macrker, et al.,1981. !
l 8-3
APPENDIX A Imad-Time Records for Charpy Specimen Tests
[
l l
1 1
l l
l I
A-0
Curw 784472-AF13 Wg : : Wp r ,
-PM - Maximim Load u Pp - Fast Fracture Load P General u GY = Yield Load I I
y I
.3 l Pp- Fast Fracture -
u Arrest Load I
> l i 1 1 I I I I I I I I I I I I I '
1 1 I I I I I I I
- et GY
- tM :
- tp Time W, = Fracture initiation Region t GY = me to GeneralMng Wp - Fracture Propagation Region t M
= me to Maximum Load tp - Time to Fast (Brittle) Fracture Start Fig. A-1-Idealized load-time record
CIBabMDC PDec a U- CL2 LIMG
. , a
~
k-T 3 q_ t n
V q_
4_ _ _ -
- e. .
.o 1.a a.4 3.6 4.o 6.o TIfE ( MSEC >'
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL2 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE careincHE poic a u- cLt urc
> i j
IT e._
A 51-n k v
q_
i l_ - - - - _ _ _ _
In 1.z' a.4' 3.6' 4.s' 6.o TIfE ( PtSCC )
2 COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL11 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-2. Load-time records for Specimens CL2 (-75 F) and CL11 (-50*F).
A-2
l CIptAMDC PDu( 2 U- CL10 LDC
, a s
. . i
.e
'l 1
g- .
u e
a l 5 n*- '
~
v
- l
~ c
_ e . , _ _ _- - -
2.4' 3. 6 ' 4.s' 6.0
,3 1.2' TIfE ( MBEC > ,
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL10 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE ennnMac posc a u- Cus use
- i I,4- +
=
^
S n*-
w 1
> w i
4 u
=
.o N_.--.1.2 a.4 s.6 4.s
- s. 0 TIfE ( MBEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL13 MATERIAL :LONG i CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE 1
Figure A-3. Load-time records for Specimens CLIO (-25'F) and CL13 ( 10'F).
1 A-3 1
.._~_.m. _ __ _ _ _ _ _ _ _ _ _ _ _ _ _
cL9 LDC
, cowcHe Post e -u- i i
, i j
"g 'g_
A
~
59-n w
i
(
- 9. -
.0
., .... .... ... .8 TIPE ( PISEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL9 i
MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE coMMCHE POuC 2 *U" CL5 USC g . .
~
~
r e
S n9- .
w
~
9-
.9 4 1.2 2.4 3.6 4.8
- 6. 0 TIPE ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL5 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-4. Load-time records for Specimens CL9 (-0*F) and CL5 (10 F).
A-4
ctpwMO C PDet a 'U" CL15 LDtG l g i . . ,
j l
+ 1 e-g a
5 *-a w
1 w_
h __
_ ~ _ _ ,
s.a a.4 3.s 4.s 6.o
.3 Tale ( MBEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL15 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE councHE pew e au- cLa umc C
E <9-3 S *n- ;
l
- w. l 2
N-
- l l -
-- - l
.o .a e.4 s.s 4.s 6.o TIPE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL3 MATERIAL :LONG f CAPSULE : COMANCHE PEAK 2 l
- "U" CAPSULE l l
l Figure A-5. Load-time records for Specimens CL15 (25'F) and CL3 (50*F).
l l
l
)
A-5
r Ct a nCHE PEAK 2 "U" CL4 LDC 9 i i
i a
1 ._
s s
n
~
S n*- +
w y_
- .0
., 1. , - ,.4 ... ...
TIME ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL4 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE
, CINNCE PEAK 2 "U" CL14 LDC ;
g i a a i r g ;_ _
s a
3 ._n w
l I
.0 1.2 2.4 .6 4.8 6. 0 TIE < MSEC >
COMANCHE PEAK 2 "U" [
SPECIMEN NUMBER :CL14 !
- MATERIAL :LONG
( CAPSULE : COMANCHE PEAK 2 :
l :"U" CAPSULE !
Figure A-6. Load-time records for Specimens CL4 (72 F) and CL14 (100 F).
I I
l e i
! A-G
i
, ethnMDE PCAk e "U* cL4 t.tMC 4 e s e e
~ ~
z A
51-n -
w w_ _
.s 1.2 E.4 3.6
. ~ 4.s 6. o TIPE ( MSEC )
COMANCHE PEAK'2 "U" SPECIMEN NUMBER :CL6 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2 l
- "U" CAPSULE !
l
. carw o E Pcmx e *u* ctr unc j
- 1. '.,_
A 39-a i
i w
y_ -
~
.s i.e' a.4' 3.s' 4.s' s. e TIPE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL7 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-7. Load-time records for Specimens CL6 (125'F) and CL7 (150*F).
A-7
, CIDeWO E PDuca "U" CLS LIMc 4 ,
I m,_
a 3 e._.
w 1
f -
.o 1.a a.4 3.s 4.s s. o TIPE ( Mste )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL8 MATERIAL :LONG CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE
, ccMancHc PEasc a "u" cLa umc 1 :-
i.. a
~
S e*-
w m_ -
.: l l
.o 1. a ' a.4' s. 6 ' 4.s' 60 T1PE ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL1 '
MATERIAL :LONG CAPSULE : COMANCHE PEAK 2 :
- "U" CAPSULE s Figure A-8. Load-time records for Specimens CL8 (200 F) and CL1 (250 F). I l
A.8 l
C!stmMCDE PEAK 2 "U" C3.12 LDIG a -
1.4 E
n S *- n w
=_
t
.D 1.2 2.4 3.6 4.s 6.0 Tite ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CL12 MATERIAL :LONG l CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE -
CtMM O E PEmk 2 "U" CT3 TIwss m.
> w I
l
\
n a(
v l v
i .e j
1 _ _
e . . .
.D 1.2 2.4 3.6 4.s 6.0 TIPE ( M3EC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT3 MATERIAL :TRANS 1
CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-9. Load-time records for Specimens CL12 (300 F) and CT3 (-125'F).
A-9
f l
I cts N NS couvor Pcm a u- , ,
e a
!.4 1 li n
". n*-
1 l
q.
I __ -
i ,
i
- 4.e 6.0 o 1.a ' a.4' 3. 6 '
T!fE C MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER : CTS MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE
, c moc etm a u- cria inws:
1.
s
~
a 3 *-n w
Tm =
y- -
1 6.0
.D 1.a a.4 3.6 4.0 TIPE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT12 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-10. Load time records for Specimens CT5 (-95 F) and CT12 (-50 F).
t
'A-10 t
CGtANCE PDec 2 *U- CTS TRANS s
g i a e g e- '
s A
3 nt- .
v u-e Y_- n n, n.
.s 1.a a.4 3.6 4.0 6.0 TIE ( MBEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT9 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE cowoc pouc a au- CTis Tanns 4
I-I l .
51- n ,
u-*
- 6. 0
.s 1.a a.4 3.6 4.s TIE ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER : CT15 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-11. Load-time records for Specimens CT9 (-25*F) and CT15 (-10 F).
A-11 F
TRANS
, CGuMCE PEAK 2 'U* CTR i i i
j i
", q_
5e e
^
$ n-9 i w
v 4.8 6.0 9 1.2 2.4 3.6 TIE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT2 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE cT Tams
, councut PEnx 2 u- i .
. i a
I q_e E
a S 9-a w
v
- f -
2.4 3.6 4.8 6.0
.9 1.2 TIE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT1 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-12. Load. time records for Specimens CT2 (0 F) and CT1 (10 F).
A-12
. (IMANDE POWC 2 U- cTil tanns
- i i e i I.
s 9-a 51- n -
w v
4 -
1
.9 1.2 2.4 3.6 4.8 6.0 TIE ( MEEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT11 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE
. councut pouc e u- cT13 tann
- i i i i l
I g-s a
51- n w
w I
n-
~
l
.5 1.2 2.4 3.6 4.8 6.0 TIE ( MBEC )
COMANCHE PEAK 2 "U" l
SPECIMEN NUMBER :CT13 I MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-13. Load-time records for Specimens CT11 (50*F) and CT13 (72*F).
A-13
, CIMAnOE PDec 2 *U* CTIO TINuts
. . . i g_*
A a
S 9-a w
lA .
I l
.c .e e.4 3. 6 4.s 6.o TIPE C HIEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT10 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE canancHc pose e u- cT7 Tams
- i I e_ _
s a
S 1-n v
y_ -
.D 1.2 2.4 3.6 4.8 6. C Titt ( MIEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT7 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-14. Load-time records for Specimens CT10 (100 F) and CT7 (125 F).
A-14
I j
I (IseocMc Ponc 2 aua cTe Tanns l J
a -I I .-
e i a
51- * \
- i l i f
2 v i
1 l
- i j 9-e . . . .
.9 1.2 2.4 3.6 4.8 6.0 TIPE ( MIEC ) l COMANCHE PEAK 2 "U" j SPECIMEN NUMBER :CT8 )
- TRANS MATERIAL
, CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE l
. convoc ronc e u- cts taans i . . .
e- -
{ *
- a a
- J 4
39-n w
l
- v. -
- 9. -
l l
= . . . .
.5 1.2 2.4 3.s 4.s 6.o l TIE ( MEEE ) !
I COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT6 l
' MATERIAL :TRANS !
CAPSULE : COMANCHE PEAK 2 )
- "U" CAPSULE l Figure A-15. Load-time records for Specimens CT8 (150 F) and CT6 (200 F).
A-15
'l i
I J
1 (XMmMDC PDtk 2 "U* CT14 TRANS e s u e
4 ,
I as_
8 i
G m
5*- n
- l 0 1.2 2.4 3.6 4.s 6.0 TIfC ( MSEC )
COMANCHE PEAK 2 "U" .
SPECIMEN NUMBER :CT14 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2 l :"U" CAPSULE l
11tmMS l , CtMANDE PDtk 2 "U* CT4 l
Ia e
a a
5 j-i w
as , - !
e . . . .
.9 1.2 2.4 3. 6 4.s 6.0 TIfE ( MSEC 3 COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CT4 MATERIAL :TRANS CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-16. Load-time records for Specimens CT14 (250*F) and CT4 (300*F).
l l A-16
i i
. CIDENtCDE PDUC e 'U* Clell 6 ELD g . . . .
I 3-7 S nt- .
w 3_ _
y_ 1 _
oe 1 h1
.9 . . . .
.s 1.s 2.4 3.6 4.s 6. o TILE ( ftSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW11 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE comncsE pouc a aua ca 8 ELE 9 . .
~
t5,n_
7
^
.f-3 ., _
~
y_
1 . _ _
I r.4' 4.8' '8
,s 1.r' 3. 6 '
TifE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CWS MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-17. Load-time records for Specimens CW11 (-125'F) and CW5 (-95'F).
A-17
, CtmmMCE PDut a "U" CW15 LE1.D g
~
I~
it S j- .
t u_ _
~
1 1
.t s.a a.4 3.s 4.s 6.o TIE ( MIEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW15 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE CW1 LELD
, ConmMCHC PDu( a 'U*
e-
"! e if n
~
S *- n w
u.
o i.a ' a.4' 3.s ' 4.s' s.o TIE ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW1 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-18. Load-time records for Specimens CW15 (-75'F) and CW1 (-60*F).
A-18 i
etManDE POhk e 'U' QM ' IELD
'l l
I 9-v 4
,a, .
^ I l
. S *- !
~
i w_* 1
.L __ _ - -
e.4 3.s 4.s 6.o
.o 1.e Trnt c narc ) l COMANCHE PEAK 2 "U" 4
SPECIMEN NUMBER :CW4 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE 1
og 6 ELD
, cuennoC Ptak e *u- , , ,
i l
I,*,~
a -
3 *a- ,
W
- s.s' 4.s' 68
, a.a ' e.4' I T!fE < f45EC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW3 ,
MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE l Figure A-19. Load-time records for Specimens CW4 (-50*F) and CW3 (-25 F). r 1
A-19
. CIMAMC>c PIM 2 'U* CW9 LE1.D g a a s a e_ _
1 1 s a
A 59- n '
w A
l
=_ _
.D 1.2 2.4 3.6 4.8 6. 0 TipE ( nstr )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW9 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE
, cowoc rim 2 u- om WELD j
" e_
s A
59" n w
w_
=_
.9 1.2 2.4 3.6 4.8 6.0 Titt ( MSEC > p COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW2 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-20. Load-time records for Specimens CW9 (-10 F) and CW2 (0 F). ,
A-20 ;
1
, muMCHE PDdC 2 'U" CW MEI.D g . . . .
I .,-
~.
m 5 e*-
w
.o 1.a a.4 s.s 4.s 6.o T:lE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW7 l
L MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE 1 I
cannot ronc e u- Cu 4 uoLo g .,_
m 5 n*-
w x-
- i
- l u_
.e s.a a.4 3.s 4.s 6.o T:tE ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW14 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-21. Load-time records for Specimens CW7 (50 F) and CW14 (72"F).
l A-21
- , ennanent rcmx a u- cus uan j
g 3-z n
59- .n w
N-
. ,. ,- ..s ... 4.. ...
TIE ( MBEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW6 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE coewcuc prax a u- cuta uns l *-
e a
51- a w
y-f o n.z' e.4' ,.s' 4.e' s.o TIE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW12 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-22. Load-time records for Specimens CW6 (100 F) and CW12 (150 F).
A-22
CIDE W C K POWC 2 *U" CM10 141.0
, a e s e 4
g;- -
z e
51-n -
. w
~
e .
a.4 s.s 4.s 6. o
.s 1.a TIPE" ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW10 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE cxpumcut pouc 2 *u* cus m.o
, i i e
j i a -
l
(
Iz e-4 a
51-M w
y-l e . .
s.0
.9 1.2 2.4 3.6 4.s TIPC ( MSEC ) )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW8 MATERIAL : WELD CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-23. Load-time records for Specimens CW10 (200 F) and CW8 (250 F).
A-23
, carvaca etas: a u- cuts sa u
_; , o e a 7
m 5 n*- -
w q- -
e s.a ' t.4' 3. 6 ' 4.s' 6. O TIK ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CW13 MATERIAL : WELD CAPSTJLE : COMANCHE PEAK 2
- "U" CAPSULE
, cowroc pram: 2 u- cHe mz
- . . . i
- g ...
7 A
5 n*- -
v
<a q- -
4
.e s.e ' a.4' 3. 6 ' 4.s' 6. o TIPE ( MstC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH9 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-24. Load-time records for Specimens CW13 (300 F) and CH9 (-225 F).
A-24
l
. CDPWMOE PfAK 2 "U" CHS MAZ g i i i i
" e_
The load-time trace results from recoil of the tup, due to the specimen not _
k 4 completely snug against the test anvils, This does not affect the reported dial y
energy.
S n9- !
v w
y_ -
{I l
e , , , ,
.0 1.2 2.4 3.6 4.8 6. 0 TIRE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH8 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE i
, ccMANCHE PEAK 2 'U* CH3 H42 )
- i i i i I
"e 1 I4 l 3
n S 9-n l
1 e
y_ _
, s_,_ _
.t 1.2 2.4 3.s 4.s s. 0 TIPE ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH3 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE l
1 Figure A-25. Load-time records for Specimens CH8 (-175 F) and CH3 (-150 F).
A-25 !
l
, coruv o c PEm a *u' cH2 MAZ 4
i-7 n
S M*-
w
.t s.e ' e.4' 3.s ' 4.s' 6.o TIPE ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH2 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE
, cssewoc PEM 2 'U* CH7 MAZ g ,
Ig_,
7 '
S *-
n v
y_ -
.L - _ _ _
"o n.a ' a.4' 3. 6 ' 4.e' 6.o TIPE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH7 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-26. Load-time records for Specimens CH2 (-125'F) and CH7 (-100 F).
A-26 l
l Cowix PEm: a u- o6 HAZ I_
z i
n !
S n9-v
=_ -
' \
i
, ,... .... .... ... ... l TIE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH5 ,
- HAZ MATERIAL
-CAPSULE : COMANCHE PEAK 2 1
- "U" CAPSULE !
l
, CD MHCHE PEM: 2 "U" CH10 H4Z l
~ ~
v z
a S 9-n w
.0 1.2 2.4 3. . . 4.8 .0 TIE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH10 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A-27. Load time records for Specimens CH5 (-75 F) and CH10 (-60 F).
A-27 i
HAZ CtMANCE PORK 2 "U* CHII e
, i a
e 2
a _
S *n-w v
y l
l _- -_
6.0
= ,
2.4 3.6 4.8
.D 1.2 T!E ( MSEC >
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CHil MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE CH4 H42
, ConanCHE PEAK 2 "U* . . .
{e
~
w I w
l -
= , T ,
4.3 6.0
.9 1.2 2.4 3.6 TIE ( MSEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH4 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE Figure A.28. Load-time records for Specimens CH11 (.50 F) and CH4 (.25'F).
A.28
-. . . . _ .. . . - . . . -. . . ~ . . . - . .. - .- - ._
l t'
j f- .. cow 80C PDUC 2 "U- Chit - Hnz g i a e s 9e -
5e e
f a
)
. S a1-w w - :
I l
9-
)
~ :
.9 1.2 2.4 3.6 4.8 6.0 TIfE C MIEC ) l COMANCHE PEAK 2 "U" I SPECIMEN NUMBER :CH12 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE couroE posc a u- cH 4 Mnz
. . . s g
I l
e,-
=
a I
A 51- n W I i
q I
l 2.4 3.6 4.8 6. 0 l
.9 1.2 TIfE ( MIEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH14 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE i l
Figure A-29. Load. time records for Specimens CH12 (0*F) and CH14 (25*F). :
i i
A-29
i i j
CtputMCE PDE 2"U* 0415 Im2
, i i : a j
I.'
e n'
l 59-n
" l r
(
I , ,
~
E.4 3.6 4.8 6.0 4 .9 1.2 '
TIE ( MEEC )
- COMANCHE PEAK 2 "U" SPECIMEN .40MBER :CH15 i
MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE d
. CIPetMC E PD K 2 "U" CHIS HAZ
- n I_._4 a
51- n w
u - i I
. , , , 1
.c .a a.4 3.6 4.s 6. o TIE ( M5EC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH13 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2 .
- "U" CAPSULE Figure A-30. Load-time records for Specimens CH15 (72*F) and CHIS (150'F).
A-30
l i
CtptWOE PEAIC 2 *U* Del ,
a a e 4 s
+
1.
e a
5 *9 f
w w
.e. -
[9 1. 2 ' 2.4' 3.6' 4.s' 6.0 .
Tilt ( MIEC )
COMANCHE PEAK 2 "U" SPECIMEN NUMBER :CH1 MATERIAL :HAZ CAPSULE : COMANCHE PEAK 2
- "U" CAPSULE ,
1 l
i 1
)
i l
Specimen Alignment Error - Data is not valid.
1 1
i I
i
- Figure A-31. Load-time records for Specimens CH1 (200 F) and CH6 (200*F). '
i A-31
.n
- f: -
lL
. ,1 :;
. Enclosure 2 to TXX 95243-L e
EVALUATION OF PRESSURIZED THERMAL SHOCK FOR THE C0HANCHE PEAK STEAM ELECTRIC STATION (CPSES) UNIT 2 July 1995 l
L i
e
't
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - " - - - - ~ ~ ~ ^ - - ^ ~ - - - - - - - - - - - ~ ' ' - - ' - ^ ^ -