ML20206E594
| ML20206E594 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1999 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-GENERAL, NUREG-1125, NUREG-1125-V20, NUDOCS 9905050177 | |
| Download: ML20206E594 (233) | |
Text
{{#Wiki_filter:.+ - ' . ,
- s l ' {;e7.-
p.,. s
.* f <-
o i*. . *u,g:. :. , ) :
; \ . . .s
,..e',r'**4 -
;.4,, /. 's
' t..- .s.'. y . ., , ,: s , b.,c. 3;'.s,:,. *
. ;, s-
,..'...-.,(.,
rq a
..s..,'..
- . .g . , ; . . n%'. ',*o .
s.
. : p,..,;,;,, .r,. g ,n**J;a. .i-(,."/.'.,,.',
,,a g;'. . .sa, ,pq f. . ,, % , <r,.,.st , '.' g n .#.
. . , ;,,, .,.,,a s;.a, i f..",*' . ..l .'r. W ' [,< :,J,b. ,a " , -' ( ,; . ]' f% M :,Y .'s , "., 4 s ,.'.;,;.,/;.. '@ ' '. , , f, 3' , ~', ,4 ' (. w" .. .' .. #
, . ' ' [ , 3:j A f, , ' M .,,,,; f,,' j,,;g {
D* .W.N, '., , .4* .-. .. A.5 'tp .j.', ', ' 4 [. y i f.; .
. t 4 '.
. k'9/j* ,'.fC;lj'.':'My, [
- ff>.f .W;T ..%,,,: ':* J. .E:.r i. ;;.W[.:. ::: j.;] '.' .'.e %,, /
.n S.,V,',', .i. Ei' E ' ; .;';[f (.i...,.,'.P.- . <.,8r .,,,I 5;',2, ,1*.[' .::;r, .%s. E;%. .
4, - . ,..,..i,;;.,c=,. .sA ....;, 4,.e.'.....'..=*.O.,. + g . . , , .. g .
, - ; . .g .
4[*? -s. ?k c
, . '.w(g., . . .- .: M. . .g,= *- .,..tr. e. ,, .i .,v
.,.. , .. .- ,e <..y g.,.1..
.i,,. . . . . ., . , , , ;..k.i. .5^
. fs d%.:. .,,f,d.,r
,,3 .0 / S ?$
'3,.,*,.,, . .
= , ',' *
..,...!'.' .~',.y F. " . 5, ,- , ,.j' * ,). . ' ,- , * , ' . .
...3,,.'.
.I ,. ~ - ' , .. ,y .: r
- l ..
. .f_.,. ,y,.
, .. : ..-:r:a, , .,., ,.#....., ,t, . . , ~ . , . . . . , . . , , . . .
.....,:L,.,,,,,......'i.'.,r...
s,..,, . .,. . - . v
.v....c: . i y.
,y .
,,.t...,,,,;;o
. . . . ., ., .,. . . . . .. .. v.,,..,A,...,.
, ,.,......4.',
. ,~; .; >
.s a,. .,,.' ,. . . . . , , .*,c, .' t- . . . , = Y .,. s .*, v e. . ;, ,.. , ...-
9.. . o :. a: > _ . +.,.**.,,,e
. , ,' , . . , , , ,y,,. ..,t .t % 3 s , e ; ,,.: ., . -o ..;y,,.-*..O.<,
- ss a, a.,.;'..",- *
'. f.e,.n 4 I , f << .., . .' ..s * * *..-' ' ., . .- :.;;y, v, ' ,.h.. .,-. is. *E * ,Y..1'.*.'.'.h,',,','.;*.,l..,...
-t - - . er.. - .- -U$. < sg M , ,. u Y,
'h.s . ,lf - t .'. . .q.:' ,-
ll e,b:Yo . so, . :, *f . , ' ,bli.s ,rfk. . . _I,).' * -
. . .,f,. ,f. .Y.h', *.," +:k.v y * . t&' ' . ..f .$* ,
s , j', t Y , . -
, ,' l
~>
J. .
., l. . ,
.:...*~ 'g . . ,; s 9. i ;. , , . , , . .y . ; ; , ' . ? t;. .. ,*'.* -
._l ,. '_, < ';; :. lg ..' f .,{c.,, , . x. , . .: );
* ,:.,_ y , , '. .M, ; ,;- :- ' . ,,
t -
.m. ) * , v- : , ,
. i* .:' . : .+_f . .;~ s , . , ~ . , ; . ..; , j,* , &, ci.M_: ..
.1 :,~.;;'; q%-;[.lg.
;*a > .
. , ' q. ; -:, : ' _ :C %, . '
- , . G l .f Y :.(* , - Q . b. ':l :?..f 'j l' f.'. ;rp :.. ;,?. ,9:N ' . . . . -i. r*.'.y..'- w.0:, ,<%. 2. d . ,t' . 4+'.-4} '* Y X' ; ; ,' . :,h.,,, g. : ., w. ,_ :"Q,. ....iJ..,i., s3. M *,l.;I .b ?' ... %
.' Q ' Q . ; ' ., '* . y i _' -i i).v . ,l-w J: ; .?.;l '.:(&'? . *,*- . ,:
La :, . Q,
.:.'.h.I.7,' i h j f. kP
* ^% '. ~k h+ . h,.W 4#'#' ' ',, 'h):,f. ;; '['
.. \ . = 1 *, % . . ;, Jhhh'.
': ..' : . .N.h,hb 9 .r ' . : i - J . ': .. ',' ,$p,k'.' . :. .; , ' .. ' . ;' :. , . ,h;N ***
, . . / p 6- .
. ' . ,. , : . . . - C :' . ', '~.. ... .u+n.'
g;. . ' , -:,\.:
?
..
- s. 1: : -
.g . Y . . , :*- : ;. '. :. H ;. ;.,3 ,. *: ', : :, y p._*.: ' .. V ' .[ ' .;;;;,' ; 1.' .s . ' ' .' : A r-' -l ; , :,. .? ; 6- , ' f g ?.,'j 1. [ , ' Q', . . . ' ' . . . .-s.- .:2. .,:. :5., ; .y s (:*..'" .(.:,.;,*..
.u . .* : ., . .l . . " , ' , . , . " , , v:,1 @, ty n* ^ ;i .i . ,'. Q;.7 + 2. - ' , ,"a. - Z ., ' '+ %c '
*: ~. -
#. p.lyi . . 'e ,.r,, - . . . . . . '+- . , . ' *, , . , - ,... * -
. . . . , '" *; ' , !. ; .. ; ,."4 * . .
...s :A :.
.,A ' ', ,.
' , . - . a E' ' , .,. . . f
--,,*,.* ,, ll' . . , ', *
. ** ., 's.
, , , , ,'".'. .a. :,; . ' -
4"'.. ..",)...
.* " '9,,*, :. , ,
,,l,..I,.
. - . , ;. .,A. :7 . ., =l * * \ ;,., ,
- - ' ,; ', g *J
,i. .' % ' t , . _y l *;'.?
.:1 : ., =,:,.':--. 2.,;y ' , ' ; .e ' .; ;. .. .k.,,,.,'lg..
; ,y, . . . . , . . ft . , ' . , , , . . R. , s' .. -,": . d./.:/..,f . , . . . .re- .i
'. '.**;,s,u.,W,c) . '. . '
, . ~. :::f, ^ .; .e,. ... . q : '! ' . , 'yG- - (. . '_ #.fv;; ,: :., ,, ...'.l.-..'. . , : l. ,:. ,* ...........:
< , , . . ,s
.C : ,,;s, '.$...'!.
- . ;;,- ' . ~ . . :. . . m. * ? " * : '. ; J . . ..
.... . - ; ^
i' . _ . , - /! ' :-
' \ ' ,' . :;.',k,3 ' ',#,-[lu i N '.'y C.i'A,- [ 1. : i l.'?: ,",,.] ",. , .I,
.! 1
.v: ,7 ')f.:v, i[ h:,1'f. 'I', ' [: k $ '. N.f v: . . . . ~ : ' ?,.f.
i , *'..' .f l [: ; . '.a .:e.',:.f.
- 3. :
v . ,d. c. : ;. - : : V ', . ,
, .y,'. 4s :-
.i. 'h.. , '.. .,p,3 .- .' .'.. .,, . . . 'I.2.,.5 . : ? +. ..,h : : I- $ ,-
gT:;'iJ :bP,';h : :a. .'./....,'e-h' : U : .:',9; :i.W..;, C' .: . .:.%.
..,:...~
- . . . ' .i .c. . *;
- v. es..: :.; . ,. , . ; ..v. . . .a; c : ., v .. ,........-q-
- . ,. . . ]!'.a.;... ......o.'
e..
,...a. 'd.
,\ . . - .
- fe,, . s. * . p ., . .. - . -
.. .i ', .= o.. . ~: . ..*.a .. . :.y
;,; .. - e. :, q; . . ,v.4, . .,. ua:,.
* ;.-.-, . , p . . a..3,- .
,[ , , . ' ,
. i,.
' , . ' , . -,Q, , .
,.-..e.... ..,,,,.:m s
= ' .. .:,,,.+ . ... .. _.. .. , . .
s
\,.,r,.,...;;.,.'.;'..+:.,..,...,..*..
. , p}.,=.,.,5,,a.-,
g . . c u
- z,
.. ...>e. g.
pr . , , , 'O ., . .. %
.- . : .<: . ,, .i....>.
.,, 9 . . Q: :,,. . , ,' . , , ,, , ,. ,,:. .; ,:, e .* :, . . '- _; - ': ; .-, , ...; .,,,j.., ,. . . . . . ' . <..'*},,1
. , . , . . _ . .. ,. ,, . , , _ . , . , . , - ., . - ,. ..,..;. - tg.
,. . . . , - .43
,t,. ;.
s.. - < * , . . . - , . . . . . s .,. ..
;I ..e' ,. '. . :
,,,.v...s . .5. , , . -. '- : a:. . , ' .'*' '
- . V._:
.l l ' * , '
. . ,1 '. " . .
l
~-
{ , <, S l-
,. , . ,, .' ,.'; . . . , . . ,,. , .i '. ;' r s ". ) .,, *
.. . y
, - , * .. . . , * , ' +
,r y e, o , ,. ; . .-
;-,';i* , . - ,',v..
*.. ' , ' - -' q *l,,f*.
,- ,3:- "
- 1. = i " -, '-<-* * * '-, ' . *. . **.,,';'-
." - ' ')./,
,,:..', , ,. Y, ,
n .;. n * -. - *.,'..<J.
- l. l . / I , ,:
~ ,
, . , . , ,........'.m.-*.4'['...,',,.
.:.6.
.,.[.,.,... .-,.,
4..
, .-. 4. , ,, .:c.-
'. .,'.7, :'.;'~,,.I-t<;'.,',','..ig..,*.-
...\ ,,, ., ' .- ' .. ' . ,,,, r.e,.*i., 1
. . . . .: , , , . , , ....~o..
- ; - ,.,,..i..,'..a,.t, .. . . .,
- .., . . . t ., . ,- .,
l '^
....- 4',*.,. ' : s- ; ,- ; '
. . . '; ., ,'- . . . _. , ,
- i, ' i ' ;*.- '-: " :; '.* ' ,
-. *A.: *1 '. , . .:' . '*..l..',. ' L. - .;. . .. , , ::,
*'. 1. .a ....
--l.. .
. : , j , a ,.
_ . , " - -. '. -i j, '* - ....
- .h .' .' , . , t:
* '. .,...L . - ,-. .*. .. ... *. '*' ' . ; *: ,'.- -'.s .,;'^':
. , ,% l , s', q . _9 - ' . . ,:, ~ s ; . '
, 4 l
. . g ,
.. y*.. - ,.-, . . . .,<._ . . .: q,, ,- ..-;. .,; *_. . . , ' " - . . . .. ., ..
n:
..,s. ,. ., ., ., .. . . , ' .y .. ,* . , ,, . . .. . - ; .....r. : . - :. .
- '. . ,r s .. - - ., s... - : . ,. .
- ra. , ,, _,,,9,,
,,,~y..-....,9..:..,..-
. . ; __g,, g.
s _._ 3 - ..; ., ?, q _...l ; y ,_;,;,,_g:- : 'g:: ..j,, ;, , :;Q l,
. T.m. ,j [ .,; ., - .1. 3 9. . ,,..*. , . ;'.
. .^y.:..: v. ,:, '.. ..;
; Q g .:; ..,,; 'y; ,. 3. ; ,'
. 4 . ._ ,'. .
- a. .3.,. , .
c..^ . , . . ~. ...::-,,;..-.. ..:. :> - , .
..:L>..,......c. . ~ . y ,.. Y. .;; . , :-
-t .,., .
. a ..-
. . . . . . .. . , -.. : 1
- y ':: -
.. .- .. , e, : ... . ..-.. .; -. . s
'[. ' : .,
; ' ,; .,, , . :' . L [,, ., . 's' g :.
- [ .
^
. - .s
. . .n :.V :.4 ~m y
- ...'-.?',,* .: - .. ' .: .. .-
..,* ,.~: . . . . . . ..:n u ~. : - .:: . c,
? - .c, .
. ;*...:.,a; , ,: .,
;; .,; , ; c: - - . . k s. ',,...u...- ;,. * :
- ,, p
,. . . ;,: : , ~ ,: * * . -
.: ..; c,: e c ;
s.
;;s .: - ;
- r. <y y,.,'... -
,;. ~,m. .
..V.....1...
,'; u
. . ,; '. 1** K . . %. ' a. ;p ;;,1 ;..'..- ' . ~ Tn r %,s.. .i;,&.y H*y-% u:W .' . ** .y '
- (,:: .:. '.
c . : N . ' i ;;. .i .j;: : . ; . s , .:". ; .:m G i, . '.W.: .l'. ;,%,
" : "* ::w-%c#.?:;
.',,%....:".",,a "n,' ?c. l.ll.5 .; , l*'
. ;; 1;.i.~f ll
':' ' :, .. . ' ,?'";, T ' 1 *:
". :: :..- *< . :::L
( . * '. gR. ' ' ,: '
' ' . . - + . ,i. . :: ... . a- t
' ^
~
,L.s l'.,j, ,; ,:Y:: ....d .- ; ;,:/- ..,:)!' .% .j.., ~.';&., d ; '<::J.'s. *. ;Lh;'s.H* 3
- Y: E. '
,_N '.l4 . 4 s"';.i .;l L. .'. '.Yb Y: .b;". :. .lot' :: :r .':. J ,. :.y, ',f .q:.G . v.q, ,;.;. i:.t. ^.a? Q , 'W ?j. .' .:.;, _}. 'l..y::.h ' ,.'
s
' ;. .. .;. * ; _' . a . : :.:,;..: ... ... :'$ :s! '
r.z5;'w"..: -'.:vl*::;45. - %.;.;'l. j !, .) ':. 's- .:l f <. . ' e. . . * ..~.q . n ' 4.) i .3 o .q, , *. . . ..-'...,.yy.- f ':
.. .'i,:.,. ..,i,,,..- ./.,. . - -
. . , . h .
. *., -s '. - [- '.,.7
'. * ')..,,,s.. .
'T'pr ' . /-
s . r, -
, ,'Y ., ^l'.; ' . ,, ;,,._l Y.
. *. > ., 5. :, . '. ; l . %.,': ' .
_,* .','..'s:
.;p ,' ;j '_
l.., *.> ' , 1.f. c, . ;.'t, -
..** : . - - *. ;\,-. '.. .:. .'-:;y-{;
Q',. e.'...i.,.
'W,
.. ,t. y . . . . * , ;.. .
- ..4 .,.. ..,:: ,* . ,. -) ,,u.<; _ yA,,) , ,. . l&.c., , :
. . . .. . ,f..,-
j
,; '. 9....... y,y.:', W. ,x,W;," _I":)l,Q:j:.w:.j;,,'
. . , . . 4.. e
-...y s '
- i :,\,.l&,[s .,'i.a ,,
- &,.,,.'W , f '. ;Q.
; .. _ s x _. .,
. : :l. ' :;. (l **
e ty;, . - .,.t,,, . c. :
. : :. ; ;. , , ' , . . - . x :s. .s .
., *, . -'a:
*-';1'-,' :[* ,r;- ;,;-'.::.- <^
*: s >.c a.. e *,e
.s. ; -
.>.,.:'.0*l-l..:a;::. .p,% ef f,:- ., ,..e?,,., r:e.-
') y -> f. ' .,.,y;; _.,.,:...'.'.,,,.t ..'.;,- - :, ."f. , +;t. .:~
. . ' ss,,.
,qq :
*. , :. .. l;, '* ; l.- -....t . . . -r .\. ,;a_'-....~ -,-'.,r,u. .':- .' t* ;. . ,. ":. .L .. ,. .' .,l.; ;- , *:, , .e .
g:
..~ ,...,c;... -: -: , f- n. . ;. . , ' r. c."...'.,.v..'s..?...
-r
* :N.4
,. '.-.,v..~ * ;: . , - ,- . . ' . . . ,< , .. .e
,'.';..:.<so'...,.'...?. - . . . ' :l .
v.~.>,..,.g... : ,
- i >
-'., : . : +,":,.
,. y o.c' .. ,. s ,s- .
- "n 9.....,.
' .r:, _ C - . . ; l. . : : :'
4, s'. ,; .S. .. ;....3 *.., o..: ,p . . v.-n. .~.,; .* : :>;. ~.'n'.
.<. ?
- s. , . g ; - . :, . 1..:n... .,._&, .. ; y,; . p.c .a.,- . ...?. : ,,(,. <'2,<
..r .
n v .y: .- : - .
^, . ' ,p
- 7 .. :;. ,
. g; ; ' ,. ._...:, ::,. _. , ... .*-;,,',., . .: ~l.~..,.a . .,- -
}
. . .'f. .;_r.*, ..r,. .e',*,,. ,.,0,.' j.n, _ n.' y 9 ,l. ,*, . . ?. .: + s .. , . . . .....,,1,. . , '. t . ; , . - 4;. s, v:?, s' .:
b -
,) ',t ,,....,.,.,;.p.i..c.,,a _ . f
.M.,.,_,V:'s'. : .?
..'...c..<.'s.. ;. , -.; .a.t:. ; 3 y, ,:... .. . . , .
- ; . s: ., :., ',. : . ;, .;v- .......sa.>,~,,.,<1
,,,....,. v. < .
1#. yc . ,e.,s* .., _, t ,n .. ;f G .p .,. ,,, . ,:. . ,,r,.. : . -.. :n; *av . -1._
- .,: ; _l, s..,q,,.._...: 1 ; ! , ;;.,, . :, ". :: . ..;; ~
.....s. . . y :l '.;.. - . .
j :., jf "':.3. .y, , :, = ,3,. :. :* ' . . . . . :f. .,.:,. . r ' ' 5 . : , , ,"..'y,. . ..::. .;":: :; q.. . . y 3 ;; ~p *-4.. 7 ., W . , y ; ., L .,.
. . .. ..a n . .; ', : . .;; , - :,, :). ' ' g-- u. : y.,; :
4
,pg ,n .:-' . . .. . y ::x : . . , ,
.,;,Q
; ,\. ;; s :' .: V . ,' :, p. -l',- :, ' **...:.
. ' . * ,A': 'a.9:, s.c. . . . ':L :. , .:::* 3.' 'I, ' ' .. l.: ; :': .. ' 'yl;. ..k.., T'i-lf: ,l' 1 :
.y- >
',.:\'.'.'.' . m;C,b wi , i.. ., .h .- ... , . : s;;?..
A . . . ' ' :g.;.' yra.
..; .., ?*., . * .:p*?.?. .;;*' . . :;. ;4 . ~.' ,. .'c, .;4 ' ;.'
- d':i . * ' . .';.,a
'r. .. ,...'.* c >q'.>, - .- *: , . : '.. . - I:,- i,::?. .*. -, ; ' ,g* ,.:,
- .:^ '. . .' O...,'M. ,
- %., h., e'
. .;- .:.-a
. . . -..:m, , ::
- :\. '.: -'h 1:' ::, ~ .
- a , s-_y:,. ?" .,e,' ' . .; ., ..,
.e. ,a .:. i..:..:, '- v i, :. .-:',.'.,, ..: ;','y .
, * . .. u . * . . : n,; w ' ',.' . * , . < . :.: s . L', .
;si .-.t -'.- ;...
.. e( 1,,&;
- r ,- l.
5.-..':,. . . , .?!, t % . y . : .'r
-g.
s a~.;<.z q: . ...h... *.* .:$'
.'::*.- '. ;1- {- l.' .' . *-4,, ,g~.'!' . * *:- , *
- i,' "'t ': t :. '*l..y {. 'n, 1^ . ' . .-
' '.Y g' ., .
p.* , . ': O e :. : : -
?.s Y , ' ; b' . ?!. ,;'.\*'
v..
.- . ,W.
.F Wlr,
' .e .:t l .%;~ . c' '.,U. .<? D.
;sn s..
?
&* <.< I./e .' . + ~-. -'
,t *:s _i . g ; , y,.. s*..
"l':'. .! &. . ' f ,1f ,. ' .N t N. .VN ,j, 'h'..;1 0*[-ll '.. . * -*.C
- . 'i.'- .. u u. ...s,.
. 4 *
.+..n. 32. . .- . . . . %* ;
- b. -h. (.*k- ".,: . sy, *.'t ,{'; ?. :*..^:;'.,i ,^[e j h': :, ,'. ,.,1,'; .:', .:l. .*l,'. . ' ': d , A. E. Y D ',;
' , .'.e*
..*b!
{. ' i ' , l ::? . k. ...,;,..ul .f;, . ,.,%.'.,.;
' S.
,'. . askl,; h". )A Q, V'.:::!.. q*; re NE'.$;b ',
. ',>~s-:
^
. i' Y &,
. o
. , ' . - R. :a)*':;r. .' '.,. : ',f~,._
-. s*: '. G.n ;,
- c. .y'.. A; , i,;s :, m s .'"n
.?,,g....,...-.. ..% *. .' ;. . _'
;.4. : *- ,
,'{*.'_'7..,,,'
- i, .o, \ ,
- .n. ,3. . i -.: a,.,
.ne..-
,;..,.y., : :
, y e -.e...';, .n A., , .:
...m.
y, . ,. q ;.lM;.y+,
.r:s .. ; *
- s.
e."g
. e e
,i .
+
- .Q .6..%..,1....G;,,>,,.s.>g*..
i,..- * :: , c
$. '..,.,..y;, r s,ys"-,,.'*.A.- o f.;g< ,.
'.3,v, ' ,; ' < , ' t. .,.x~~,,,.,,.1.r'.,s..,,...v.
.. . b _ .; .s. . :. . ., ,pg . :,; q. s- .
t..
, * ^g.e s * .J',.'
~.1
... c,,:. +; %
- .\;3: . ,'. j ,~j y-~.j,.e
. ,. .. .. . . , - e;, . . .' , . ' .
., s.
a
..c ,s a . ., q . _,,.s.
- ph....>,,...
.s , :. :
. ..< ..' 7..: .;,
. . s
. ;9.
y t ..
- t . . ' . ' ..
, . ),: ' ' ;4
.'.'y:.N.,: , R +I. *
- i'a. ;; . .-'.f,2 ,f., : ( s {;.}.'a c',,-
q ,.,a:,:q;l'...' , '.,4..* .,.3..<.s. : . ? ;: .{.1, 'W L. ,.n)% . f. 'at -
- t. \, lv s Y, ; ;' ,.^', f 'gh.: , yY. . i ,h',('d . lo . ' L'. i:( i'.: s'd ' ,:!l le. 6, q:*; . .
.. ,*, *.. , . .;. !. ::.v- ::p :n v:. ,l',- -. .',',e':..
- L , p : . r s,,,.._. . . . :: c - ..c ,L . ;.- t - ,.s..
.,-;' ;.6. .";. 7. ;. y , '. .. .-':' ,.':
s
.. ; .r. .
> r ,' _;. . :, . ~ # . m. g ;, ....**;...;,;_s ; 3, . r .y' . ;,,n... .; .:
;: :;. y. ,, g.*.n;. *.....:. :. 4:Q. , ,e.n ,.g.-,.,, ;; :
- v. .,.r. , * :.;.
s e.
' . s :. ,,l.... . , ' , '. ':. ... 's . ;. ;3 :; . .u.. . .; . y. . :. : .,. . s; , ; e ,- .p.._% <,..:... ,, q ,;..'.
;:l ..-;:. .:e,~... *, .,..)...<:'y "; s
-Q: ._r,.
. . . . .. . . . . , .;, ;a, ,,F*,,',.,,:.,.,.,
.,v- > ,.. ; . . .-(. f,..1.1...,.* ;,.....',>-. . .
x,,,~., ,.......e.4,..s, "9,'1 .
, > -: r. ..0
. ..r. -:, . s ... ~ . . .. ,; _. , , ., , , .. .
- *) <l' f. .,,' . -; ;f.s,.r . . '. '.. .'. -l f,. , . -\s ' ,, .h ". .
;, .y Y.
. '.:~?
1, '. .'.l.l} ..
. . ,: _f...': W.
- r , .- ' c. , ., $ l >; s $;. .,; e.
* ;, .v.c.. .[',."..
,l , !
fl:'ll',, ,. ' . I:f.:,:', .l. l; ', .y .? ' '. ? '., .,,Y,r
- - - ,., ,..-: , : :..: .?S ; : , g .7, l . ,l
.:'... , ,. - - ~- ,;.'t,, ' - .yc - , : . . .
, -' m v .
. r .
~ . ," : , ,, . s. y , " . . s. < . %,. z ,: ' , : :
- , . . . - . J r--
e.
- , ; ~.
.., s
.p; . - -- - .:
e
^. i. ,.,s.,u 'c:(n,s. * :, , ;, . -, .
.[ . - l:; ;. :.'.,, k';'d b;' # : ,' i,7 .'6 h; . , . p.:. .;.
t
';y .* .
;, ] 'f , . ', y ,;- f ', , ; '/, p ' ., ; .,:'- .,' . , ,:. .--
e 9~ * :.4,v.1,,,..,.,,., .p,,,,,, <-,p
. ,M, . r -
t.'l.,i,. .,, ..s ,s-;i,-':.,.*
,.s.. - , . . , ,
.n
- p., . . , ' .; . ', .,
; 4- ..-, 'q ., '..,,,.- . o .,;8,, ,,,,, ". f.,f.;.l; ~^ + .,,f. :..,.;,.., ,.s E . ,.' ,. . -*':,-
C. .: .;. , -. ,s *'.g: 1.,..--.. ,,...g',..- ;, ,,y,p,tv- ..,.%,.>.,;},.,'..p.',
" ,g . ),. , . , , ,.,.L.
t f;; . .' ; t ,,,.,:.,.*;..,- ': ),' [ ; l.' O. ' f., &., ,',; (;_O,
.-^ ,,,g..1 , . , . . .. ,. .. ' . . . . . . ,4,,,
c
. 3. . , ,
j ;;g or.:, .' ' , , , ' .
.':_ .!,'., 1 -' ,\ ; ? . . ', ', .} .,l g ,.y_,rl ': d -
j d'*i ' W
, , ;*.*'; . . _.l .:+ :j ,._". > l:,'^ ..;['.[g. ',:.. . . f , i . ?) ,;':".[.~;. '{ ,r,~. j
,$j,, , . .,, ....,,._;.,., p;.;{,. ,'.f,.
l,,
.: -. ,(,,%. .. : ;
. . . -.. ,:. . . .., ,> .,. m,>: :- , . -_ . ,:.. ...p:....-,.
; g.,.. .
..g,.:.,' .. : < a. ,., ,,c... ,_ :, s._
i,.e
...,t,-v v . .
_ . l. : . .
).
- _ . > ;. r ,..::,,.u.,.,..
- s. . , . . , . . . . . .., :. ,:. r ..,i.,.-.., .,.,,...,: , ..* , . , ., . .. .. . ...
.. .. . a . ' , .s ,.., ~' ,.~. .:. j... .-, ,e- s . . . . . ,.p.t........e,. . . . . , ,,..
, . , . . , .,a.....t;<.. :...+. . .: , . . ..
- .. -),..... ,... . . . . <,s,K...~.;*...+';... g, ,.*m..',."...
.......,.g[, ',.,;7..... . . . .- . . .:: . . ',. . .s. .
, . , , .. . } .:. .;..,. .e..
,.,...q..:n,.;....s..
,'.- L.s % l sy.:'
1.,.$.'.':,,. .
. 'i :.,.,t / g, {, -;. ,'. ; , '}:,
.' 7 rp '.' ;
- l i,, .(
.j ., '[l' [' { .< ! ;? , . j_ . p,: ,' ., , . :g..(,.,t'..,,<..j+...
..,. , . ...s..,.
{z { , . g /p. . .y :. . ,' . ' ,',4
,s
. r..-,, .- ,-{,-,, . j.,.s - e..-)....*p*",
'%;4' '.
- .f::': :. L ' . . .:]f . : :.:: . -Q,.< n:
- Q ' %,.i i 't c , .' - " L . '.= T;. ' . %. . ;' e::Li b . * . .; ' = :., ': : ,. ', 3.? : . & v&:: . . ~M '. .; ,:: C' . .l:: : :4 ',~ .: . : . :. . . 1 : .N; ..~; '.w .
. . . > .. m. ' , ' - l. Y . . '
e_ 4
.;;7 ,, ' * * .', i : * ', . 2, ; . p( , , ." ., , rk . ; e l.,. ' . , ,,,. , : ,
- s. f .:n .
,,,.s.:>-,
- , < ,. ; .. ,<: i[.., ; - s.
,..,,16 ., _ . .::_. .. ;,.,...' ..c-, )_ 4*,,,..- _ - . ;; .; ,". .,g- . ,, t j . ' . t:-l.,..,
k-
; ' . .' . - - ,' ,', :. ,' - l. ? : ;; ,,I:. .
Yl: ,. d ' . [; .,, ' ' *:' h .'i, ". ' .\. . ',,,'.s'
' $ [ '
,,,','(t.r.,.
sr ..Q. y , g.., ,. :h . , ; *. ,. ;.< : . ;;ys
-,4
,[,..1,j *, ' d'[ ._
.[ .;~v.[ f.L...k. , ',;ll[.,
V . ., 5 E' 2
> *,['...
...:.r.._,
. . .' , 9
# E ..'
M'Q(._.."., . ,- . .o,... , .) s ;s
,e a,.
). :. -
f, n'h.;;:. .
,., pg
. . . ., N
. ,.y-. . .
*. ... .,'.-3.f.1 :;~.r; ; ys ,4O. ;qonmt;~~,,,..s:t.v. .,, ,Q. ; .,l ~r *,7.
G( ,. _ , ,, . : ,:..,, ll - , \.r,-_..,::.y . _ p.g--.,[:-, q ,.,:. x: y * , .1 :...,;'....<.v.,'.;c: p t.z.+.,,*.,w
.H l>,e,;,,>','_
.- : (.: /* ) ,,;. . .. v .. . . : ; 9 . ' ?- n ' - ~
a ,: :
* :y(.'.:*>,',r .-% .' s . * , ? p;...'.. ...
q ..;. f. . :". .; *. 3; .r):.-::y , .. . ;.y. g. ' , s , . . g y , .l L'g:;. s ;.
.,...,c'.,s.,;
4,
- -l
'E
- . _. . : . :.. c.:
,o ..
s .:.:..
,9
.s..,.s...,. . ;:';; g W u,
p , .3 %-m; v i., .-.. s . .+ Ly_. .:.- \',.;'5 ( , . .i
,;; p
, r ::...
n y, .,
.a ..e
, ./. .,:$ t
,y:,;,.';lt
_ ,~. . c3; .. .. s.x- ,s..s'w ., . c ." ' .
- .,r...;-, y
- r. .!*.y g e.
,.,s'....,..lo.r* - .......:.sm."'tg..y,.
...~ .; ,. s e y: :s ;. . ..
.( . - m: . ..::
t.r
.c ;. #, .
4,
. . ...ux . : ,,. ; _
r
.. - - : . . :. . b. .l.a -
AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications NRC publications in the NUREG series, NRC regu- NRC Public Document Room lations, and Title 10, Energy, of the Code of Federal 2120 L Street, N.W., Lower Level Regulations, may be purchased from one of the fol- Washington, DC 20555-0001 lowing sources: < http://www.nrc. gov /NRC/PD R/pdr1.htm >
- 1. The Superintendent of Documents U.S. Government Printing Office Microfiche of most NRC documents made publicly RO. Box 37082 available since January 1981 may be found in the Washington, DC 20402-9328 Local Public Document Rooms (LPDRs) located in
< http://www. access. g po. gov /s u_ docs > the vicinity of nuclear power plants. The locations 202 - 512 -1800 of the LPDRs may be obtained from the PDR (see previ us paragraph) or through:
- 2. The National Technical Information Service Springfield, VA 22161 -0002 <http://www.nrc. gov /NRC/NUREGS/
<http://www.ntis. gov /ordernow> SR1350/V9/lpdr/html>
703 -487- 4650 Publicly released documents include, to name a The NUREG series comprises (1) brochures few, NUREG-series reports; Federal Register no-(NUREG/BR-XXXX), (2) proceedings of confer- tices; applicant, licensee, and vendor documents ences (NUREG/CP-XXXX), (3) reports resulting and correspondence; NRC correspondence and from international agreements (NUREG/lA-XXXX), internal memoranda; bulletins and information no-(4) technical and administrative reports and books tices; inspection and investigation reports; licens- [(NUREG-XXXX) or (NUREG/CR-XXXX)], and (5) ee event reports; and Commission papers and compilations of legal decisions and orders of the their attachments. Commission and Atomic and Safety Licensing Boards and of Office Directors' decisions under Documents available from public and special tech-Section 2.206 of NRC's regulations (NUREG. nical libraries include all open literature items, such XXXX) as books, journal articles, and transactions, Feder-al Register notices, Federal and State legislation, A single copy of each NRC draft report is available and congressional reports. Such documents as free, to the extent of supply, upon written request theses, dissertations, foreign reports and transla-as follows: tions, and non-NRC conference proceedings may Address: Office of the Chief Information Officer be purchased from their sponsoring organization. Reproduction and Distribution Copies of industry codes and standards used in a Services Section substantive manner in the NRC regulatory process U.S. Nuclear Regulatory Commission are maintained at the NRC Library, Two White Flint Washington, DC 20555-0001 North, 11545 Rockville Pike, Rockville, MD E-mail: < DISTRIBUTION @nrc. gov > 20852-2738. These standards are available in the Facsimile: 301 -415 -2289 library for reference use by the public Codes and A portion of NRC regulatory and technicalinforma- standards are usually copyrighted and may be tion is available at NRC's World Wide Web site: purchased from the originating organizat,on i or, if they are American National Standards, from-
< http://www.nrc. gov >
American National Standards Institute All NRC documents released to the public are avail. 11 West 42nd Street able for inspection or copying for a fee, in. paper, New York, NY 10036-8002 microfiche, or, in some cases, diskette, from the < http://www. ansi.org > Public Document Room (PDR): 212- 642-4900
O:' i . .,4
;,IU,Y,' (( ; b. .
b ' .51;N ;d . '; ,.l."..PJ '; /. i .j .. '? '
',2 .,.l,.<{ ,v e-{ { S I p 4,4;i af ,} ;,y b %g/' , , . ),[ ~tJ ?. i. ' . ..',r. i,$ y,#. Q;$ . . i) , ,., JI c,'i.)' .
,' .: m
.[.5, ' ,.;., . ., f.k ;1 'p k.%[f . ' , ,'d , . il]} ,.,rf
.y'" . : 38,,l.N. 'V[; .(j '
,5,J, . l. '. . ., :. * , '
'l [ f,' ), h,f,hh.[: _, . ' . .,
,2;E (N. , ., , y
. k'^
; , y ,' ,.T , .-[,1 {.k,h'M# .hl.,, , .:. :[ [, ,[, , *
, l .
~
^
, ,. a' ,. '. i * .
','I, : :
l *-<:.'4, i,f .
.....- s. s.. !;i . I ,- . ;. :<_'. gt ,
- e. !. . ". ._r.,.
f,; 4, . . ' .;;; - ;
^L- - . . ,. r ,;.,. - s
- .y ?.' e . '
[ ,,,;,,,.?.,,'e'.
.l'.j;,,l ,.v:; l* }'. Dl .* .,.a "]j .,.. ,;, . ._. j y. *l..;',[: ,f .
E.". ,. J *.' ..; V[j,l'.;.!. ', :.&a , 3.&, _ -.;-* . .i' .;'~ .' i' !_}. ' ' y'.j. .[F: '.:'.. ',; ,} .,.;;:} _%. . 3,. i7 -:J i } 'l y ; j.._.' f. j: '.?,? }'; A.- 'W
,.v.;.,,,,, ,c..ly> 3j:.?,f,:.
*I . :b,,
+ ' '.'.p. as .C*:.
-: . . . . ' 7
., .'.l p.;* .
.*,'i'.',,,.',..),.., g .,it *
'. ,.: ,k .*~ , . .
, . . - ;)
'> a . '^
u
/
' **;' ' - T - /* ,'-.y ;df i-' '.f j ; ... . , ' 'g'-
- , ,.: v: :. " . .: s -',,. - :. x i, , m T, ' X _* : '.A. ' : ,^." b ? . . '. ' 9'.
, e ._u. ;.- ... ,,. .. ., _. _,:--, .,.;, ,j, ,.,, -
-. .. .. ..- . , . . . . t.'..
.;.,<:,'.'..,.'.,1.n'...;..;s : -'- ;Y.2
. ' .;, s - ' ;.,ji..
q'...;y.....,. 1.4. '. . a : ; , . .,, ; , . _:4.,, ,;,,,',,' : .:.
. _..',.,..;_q. 'y '.h'.''~_.~n
.s, ,..'.. ,
. . . ' , ,, , , - 6 -
s'
- =
1
.; 7 '.' t , :,; %
. , . . ., . .; . . . . . ,. ,. , . ,.,.._..,o.,.
g
..v...
y
. c, . ' . .
4 I. , *. .
- z
,<.. s...... ..,. ., , . . .. , .
,, , . _ . ,y.
_..g .:. . . .:~,,,;,._p , ;. ...,' .. .. . :~..
.. ~ .. .
. _ . . ... ~ .. . _ , ...
5,.. .. _ _ _, r . y
- ), . f
.h.#. , . . ..: .. , . ... - ,...; .~ n .. .. - .. .. . . . . . .
. s .o .
.... .-.. - . a - . ; c. . . .- , ... - . . . 1 . ..,
.,v,,.. 'e- -
,' . 3
, . -: . ,- , , * ,: 2 , . . . .- ~ '
y:., ..yy .
.,r- .: j f .f 7.: - . . - " " ..:-.
. '. , s. 3.. - - : - - - -
( __ . 3 . .: . - '- ,, : . __ . . , . . _, . . , . . 3 .
, ., .- ._ ; ,, l <. . ._
,.y,_,
_...- .- - .. ,_ .. . . . r . k . , , ;- _' . '. * : ^~ j :. . : ;;. . ;, .c. .' - v., .-(., ,,. ._;.- .g.. > . ' . .., ._ ~ , y.'
- . :. . ,~.- , - ..
, . , . _- .. . - _ . : l..,_ _ .> , .< 'e _., ;.: .. . ;,: . , >
.;,,g.-
.- ,: ..;.- _ . . . . . ,;, . - . y'.. >,~...,._._.,._ s q:.. ,y .;: . ; 2. - Q : J - '.i . , , :, . . .. .....,.;..
, . -w.
. , . ~ . . . . . ,. ' y :. f. . ; . _. l Xf. W..:. . . .
. .. : ; : . :,' . ' : e?.
..-_~.,-__
.~ .. ,.
. " . . . . . . . ' . . .f. .
,7 5 . , _ .,.,,....;. ..:. y.
. ~ ,
r
.~[
** : ? i y : :. . ;j. : U;' :
e ' <- '
.: ,s. : : ,,, j
.; .,;ll_..-?..'...',- ~.. , ). :,c.
.....h.,,
- :..'.. . ;. .' f , . - ']
s.' .c....'-* ..,, - y"- ,. , . Q . - ;- . . , . . * . > : -
., f . . m '. ': ) :.. ..'l . ' . . ... . ,~,,..;- . '.. .
- ._3...' .. . .;3; .'. ;. ., '. ., . .o._. '-; -m _...;r_,.... . . . :: ..: .. . . -
.i.y
'4 - ' ~ . .e.,
- - '.:, . . ., ; ,. , :. ..3 .. r. '.:< - .. . . ': _ ='
,n:, ... - '. ,. .. ' --. . .l: 'f.
: s ;-
-)...<. ..
- Y '; . , ~_ . . . .
' . (..
. * . .y ,i. .:'j ,
.' <'.'._.;,..:... ;.n.-
.- :,-e . :. ' ., :' :
. ,...',.:- '. ,_,..,, . , _;, , y ' , 'r- .,.. e. ,.;.'_ . . . , .
':. . . - . . . . _., _ ' . , -,_ ; :- . _. 9-n *
- ( _ - .y ,. ..,.. ;_ ...
- . ;
p;.*...'..,f.,, : e
.p. ,.:.'..r.s'.- . . , . . . _ .
* .. ,, . ;,- i , . , ;y .
.,..ll. . . . -.;.. /-' ..c..,... ..
. w ... - . , , , . .. sv . ,.,. , .
.,,b.., ,., .i<
+,7 k; a ,;;', , .-' _' ;.,
, . .t'.
.! ' ( . , }' . ' ,'-' ' ' ~.'. ' ; - I ". ' +. , ' .f j ,'6 -
; . ,' 'W h ,. . : -
. < V,'.r,'.'. ~.}..'.;;...,'.,"'."
,,, ' Yl...- , - # ;-, ' _
.; .. m;^'.,;..: , . , .,. ;, . y :, .4 *. - . .: $. . . :
- y
- ' :. . , ' :. . ,b ' ~ ,. L a '*: ',*l.' . l, ; . . ' J.. )' , . . .'
s
. ' ' . T;. '. . ' -
', :: '. .. ...".:,l~Ts--,
;* ;_, _,__,._..p - - '
* .{ ,l*. ,R. . . . ,-[ }r.. .', .'.,: .? ,
7 , x. . _., _ . ' s ;> [.'.,,.l.
, _. ... . . ' - .. ..,,,~. ;. ,
-- : ;'..VE,_
i
. ; _: -'.:', ; ', _,....,';> . .} ' l
. , _ . , . , , ..___. 7
, ~ . : .. ._ .;.g ; . * , rl ., , , . > ,, , ' . , 'q, - Y.* .
" _ _ . .,. } _ ,,'.,.}>,_.:
- . . .' .,s. ?-a .,.,m. . q& .:: , , ._
. . [,
. ._ .'. ,-, ..;- . . . ,.. ,x_ . . . , . .
f f' f ';. . ,. j: p, . .et . : .,.. v. .,,._,:.- ,. ._
. _ . , ' , . . . '. ',y
, ' . ,. ' \ .: . y 'g : -, ,. -' . , ' . y ,a . . ,f,,
s .. , , , 4,P
. . . , . ,,.,f'-
F
. . . . . 'a .
j; ', g ,,' . d ' '
.. . ' - .., ^ ', " . . ...l'j,,. *:,,
1l ':? '. l, . .:j '
,'+ . : p . ,, , . . :.. , . , _. .j
; ' , * . f G,.: ]: ' i ;'. '.. '..] ' .s, . ';**
. . . , . , . _ . ~ ;r . . . ;- . '. ; ;
. ' , ,.\. . ( ! .i. . :. ',. f. ' ; ' l .: .', '; :,' ~. ':' : : .*., . :.; ; f.: .,, * ' f.. ., -
- ' :. . ' . '.:'%..s..' f i; .. ' .Y -' - ':%p"
. .,,;,,-.. ...'. .'.. , . , . . , c
;m
...s,,'...._. ... L'.' , ..
-A.* ..;,* , . . . _ ,. ... * ., . .
,s.- *; . , ,,
~; [
t i.
.f
[ ,', _f. ~*;s';',, s '- l;
.,;. .'.' .;;[ ,;.,_;n
; .g y 9,;,;- -'y'..[..',.4,.'
f .1,f . .. +. ;c,. p
, . . - - ;j_
's
,_:. 4.
i ' ' , . " ' . '. '.' ,. . . , .*,,ec.,*+- o
,,,.' ,,._.,.,j,,lt,'.
' ,..a-,',y.t-
-^
.9
. ;., ,.'.v. ';:
, n;
. . . 3* '-; , ' .; ; ; ,, _ '
f._,k .. . . . ;, .L. ._[.;'.,.;' _ . .,i.. _;_. ., ._ ., ', : s .1;, . h*y
'. ,j;j, .5 l
,J. *-
. :. 4 ,.N?'.', : 3. ', , . i .', .' ' ]*.
- t_ ' -
, , . . . - - y,; ,',t- :: : " . .' , ';' , .
.,f ., .-
;.,;.'J- ' .' ' .' , . . . ,,4
.'l, - ' '
g t. 4, M. '.g .c, Jr r.?.t".., c.:.: ,'
.-' . .' ? ', ' '.; , c. ',, ;....'.,L ".../
gl ,p #. ','
.: / M ,. :. ,,' 6 .._"_.",j....;,(...- *',f' c%. g,,).-f,.. . f .p, e, f : . T,1 . ..',I.,,.'t. * ' ,' ,. ; N,' p '
4 - r ?- '
.. j , . . f,
]f,L * ;. *. - * ' g*, , ';,} y ; , ..o , ,*.*Vs': , ;.:, ,;.. j '.; . ' ( ' ' . q , ,y _ .%] - ',,..[:. . , ,}. ; .;e . ;j . . . - .
, . , .s
. . Q . 4. .?.*. . :: i. s ^. ..-8
*. ' . ' :',.. ,.: .'. i.. : , , ,* . z ' i?..j [ ' ,' _4 .,, .,
* , - _ ; 7: '. . y l j;
.c , 'j'- '....:1,.,.4.'
,; ,[: ,7 ',,.[..*;. _ [, ..- '
- 2_'
.,,...f_..,..
,i.'..' . g
/ . .
3.',-.,-;;y'. ' : ., , { ','. . .- . ,,,e..?*_e.
. . e.: i '
'; . : ..' .'.,G
, 'c 's :,,. . ',:,.
' ...;. o, ,c. . , h. ,f *, .': oW~ * '..?,,.{. -' . .. . .-...,j "<<.....s..,.- ._ ..
, , .. '. Y [ , 8 ' ,: ;-
. . .r. ,sj ..- . . . _ . . , . .e..,.. "".,'f'
.,n'
-. w '. ' . , . ;. y ....,;.'l.';,',;...,
-l '
, . . . .,,;, ,- ?. .. j
./e -
l
,.g ,.
ff ; g, <
-l ."l . fJ - , . . , , .* ' ' - ,,
y.'.'.;.."..
~y. .,.'_?.
.g,
;. .;. ._ . . ' .,. ,. . , .: " . - , ..',.,,,,,.,6,..
- ..h..'..<...-
+s
',,'g,
.;.. e.'
*'%..,, . ' , . ' , ,6,'.'-.
- . , .*/ .1 '._-
- .. . $' .' . . : ," b
- j. ,
'[
.,;a. -
.._[,.,
+: i, -. ,. , ~ .
3
',_ _~ ~'_~'._ :..-l_, .:_- ,
-', , ' ,. . g ; ., .-, ;., ,.';..t_
, . . . ,N
.I
,',g, %
, ', . :[ : . , -' ' ,L, , . ' , : .' , ' .
~;:. "i
,.'N.,...'_. .-j- .,. 9 l * , ; ' e .. ,' ^;'_
. . , ~ _ * , 0, ; ' , ' . . . . 3 ;,,, _':'. .{. ' . ,.'..'.Y-
. . ,, -[ ., ;, ; . ' ' , -
.f.
- i.; , .p.,'...'..s(
y-i _ a,,e e 2. .
. , j . . ,
. , ::-'. . , o . ; : , ,;' -t. . . . . ..,'1-- - ' ; .- . 3:, p:;; ';,,,j, .{ .;.. ... ,7- _ .':
. e;
, . -,.'. .-.'s". ; . .' l; p : ,, . ..lt._.,
- s. . ,,
~ '
- - , .g, i,
.. .. 7 ;, .,
', [ , ..' . ) , .'.s '
l~' ',
;}
- U -l' ' -' ;, }_' '
'!y ' --..
, ) ' ,; $[ * * $,- i . . . . . , ,
.[
; ' ;,; . >, . . ,._ . . ., " , *'_: ..._;. , ' , .f. . ' ... ', ., , - ', .. -. ' [: ,' , .. ;[gi . '.- '[.I*'.,.
T ' Y, '...L,*' ; . . ..). * . : ' ' ' ,,; ..: . . . .:-
.::-* . " .'_s -', 's...j,.;r. =;;_, , , .' . ." . _ ,,'s. .. o'"_ . -n.. _4. .o .- : .'.g
._; . . 2 .1 ;L.h - -
Z 9
, ;;_._1*,.. ._,_.. < ,. . .
;- . *,i , ;: -
v t F g. t,'*'l,,, ,j';'. .l * ' k ,l::),. , . , .-;
. _ . , '. -;..' . . . ..':J',
. . . . :: .; , :;=,l,: ^?- T' r.*h'.,
.c . , .-
-..;. - - Q. ' g . l-' ] 's - '* ' ' ' . ' ;, . .'l,- r ~. .-,-_ . ;
a'
~ _ . . * ',.- ..';.'?*' " ~ . . .f : "* : ' f. ~ :.; ?
'. .. .'.;..- , .V . , . _' r e,': ::. ' . . :~ . . .- M. , ,_. .[ ? , . ,; ._ . _
,.3 :.. _s.
7','f..,
* '- ' .'. 6 : . . . ,'- . . . . * . . .
'____J.'.. , . s
' .- ',.' ,.: g*
e
- A . _.:
' . ' , ; * ..;. -;. -,: . 1 s_' .v'
. . - ' ' :n ; ;'x.,.r*t..-*/ g'..
- g> - - ,.. ., ' ., .' ....,*
,..t , .
., .s,. ;
.'p. ..n : , K.. . . .+:.,.'.<*...,.. . ::z.' g. ;s ..i.:.,,'..,l.,.,..:\
. , -. . . . -. ,a . . . - .,..
y q,' r 4., i. :'.. r .. : ...,;
' .'d :' . : . . '. .t ; .'", . " . ' . . ~. 1..:.. ..
...x.--.
...s s . c f :u':
v -'...:.' .,nn :x.
' . n
- . .n ..: .: . > . q .. ..':,,,
L.. 7* - ':'J.:;_,, . . . . ; . l.' , _ ~ . . ~ ..- : .- .
- t. : ;, ; ,c ) ,9 .. ;yi A. <,;..:':., ;,Q .xt,. . ; : . : . . . . ',. n .. *.; :~. ' . . ..~ :f..r.:
...r; : .
7.-;/ /
.! -4....;- . . . . . < '.. . n . ? ; ;: :, * ; . . .:: :.;
..'4
. . e:L' ..l'... (: : ; -
.u ;.;-
- q ' , , ;. :- :: :::r ~
.. ? :.
' - . .- is
.. :: : . c. ; t,.. T g *ij - . ;
$ . : <~ y, , .2 n.c .> :. T .. l:V . <,1
- ..:; q . : . ; : . .s...
0: ; . . . ..; ^
~
. ^.: .t'. ;,,
c
- L: .a: . :l y ... n. .4 .~
- l. : .
u,.
. '- ':.H;.:.. :gy:.:: ...+ei n .m N{
. 'y l .;. . : '. G
. 'Y . : +
- + . a ' w'..<'s*. . j: .'s ,. v .. '^ '); .:a. - ' ' ';-'u .
.~q ...;., .y, . q.:.::n;y;.u s'
..~s." :-Qb f:(f v %+: b: .~;?:4 K
P.p:.,.. ,,: L. .g: wyyyypywrmwgwweenwwweexeeerowyM. l.' f
AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications NRC publications in the NUREG series, NRC regu- NRC Public Document Room lations, and Title 10, Energy, of the Code of Federal 2120 L Street, N.W., Lower Level Regulations, may be purchased from one of the fol- Washington, DC 20555-0001 , lowing sources: < http://www.nrc. gov /NRC/PDR/pdr1.htm >
~
- 1. The Superintendent of Documents U.S. Government Printing Office Microfiche of most NRC documents made publicly RO. Box 37082 available since January 1981 may be found in the Washington, DC 20402-9328 Local Public Document Rooms (LPDRs) located in
< http://www. access.gpo. gov /su_ docs > the vicinity of nuclear power plants.1 no locations 202-512 -1800 of the LPDRs may be obtained from the PDR (see Previ US Paragraph) or through:
- 2. The National Technical Information Service Springfield, VA 22161-0002 <http://www.nrc. gov /NRC/NUREGS/
<http://www.ntis. gov /ordernow> SR1350/V9/lpdr/html>
703-487- 4650 Publicly released documents include, to name a The NUREG series comprises (1) brochures few, NUREG-series reports; Federal Register no-(NUREG/BR-XXXX), (2) proceedings of confer- tices; applicant, licensee, and vendor documents ences (NUREG/CP-XXXX), (3) reports resulting and correspondence; NRC correspondence and from international agreements (NUREG/lA-XXXX), internal memoranda; bulletins and information no-(4) technical and administrative reports and books tices; inspection and investigation reports; licens- [(NUREG-XXXX) or (NUREG/CR-XXXX)), and (5) ee event reports; and Commission papers and compilations of legal decisions and orders of the their attachments. Commission and Atomic and Safety Ucensing Boards and of Office Directors' decisions under Documents available from public and special tech-Section 2.206 of NRC's regulations (NUREG. nicallibraries include all open literature items, such XXXX). as books, journal articles, and transactions, Feder- l al Register notices, Federal and State legislation, l A single copy of each NRC draft report is available and congressional reports. Such documents as free, to the extent of supply, upon written request theses, dissertations, foreign reports and transla-as follows: tions, and non-NRC conference proceedings may be purchased from their sponsoring organization. Address: Office of the Chief information Officer Reproduction and Distribution Copies of industry codes and standards used in a Services Section substantive manner in the NRC regulatory process U.S. Nuclear Regulatory Commission are maintained at the NRC Library, Two White Flint i Washington, DC 20555-0001 North, 11545 Rockville Pike, Rockville, MD l E-mail: < DISTRIBUTION @nrc. gov > 20852-2738. These standards are available in the Facsimile: 301 -415- 2289 library for reference use by the public. Codes and A portion of NRC regulatory and technicalinforma- standards are usually copyrighted and may be tion is available at NRC's World Wide Web site: purchased from the onginating organization or, if they are American National Standards, from-
<http://www.nrc. gov >
American National Standards Institute l All NRC documents released to the public are avail. 11 West 42nd Street able for inspection or copying for a fee, in, paper, New York, NY 10036-8002 microfiche, or, in some cases, diskette, from the <http://www. ansi.org > Public Document Room (PDR): 212- 642-4900
NUREG-1125 i A Volume 20 i s
)
A Compilation of Reports of The Advisory - Committee on Reactor Safeguards 1998 Annual U. S. Nuclear Regulatory l I Commission
)
DFE>2_ l April 1999 l
ABSTFACT This compilation contains 59 ACRS reports submitted to the U. S. Nuclear Regulatory Commission (NRC), or to the NRC Executive Director for Operations, during calendar year 1998. It also includes a report to the Congress on the NRC Safety Research Program. In addition, a report to the Commission on the NRC Safety Research Program, NUREG-1625, Volume 1, is included by reference only. All reports have been made available to the public through the NRC Public Document Room, the U. S. Library of Congress, and the Internet at http://www.nrc. gov /ACRSACNW. The reports are organized in chronological order. iii
r PREFACE The enclosed reports represent the recommendations and comments of the U.' S. NRC's Advisory Committee on Reactor Safeguards during calendar year 1998. NUREG-1125 is published annually. Previous issues are as follows: Volume Inclusive Dates 1 through 6 September 1957 through December 1984 7 Calendar Year 1985 8 Calendar Year 1986 9 Calendar Year 1987 10 Calendar Year 1988 11 Calendar Year 1989 12 Calendar Year 1990 13 Calendar Year 1991 14 Calendar Year 1992 15 Calendar Year 1993 16 Calendar Year 1994 17 Calendar Year 1995 18 Calendar Year 1996 19 Calendar Year 1997 l l l V
F l l l
~
l ACRS MEMBERSHIP (1998) l CHAIRMAN: Dr. Robert L. Scale, Professor Emeritus University of Arizona l l VICE CHAIRMAN: Dr. Dana A. Powers Sandia National Laboratories MEMBERS: Dr. George Apostolakis, Professor Massachusetts Institute of Technology Mr. John J. Barton, Retired GPU Nuclear Corporation Dr. Mario H. Fontana, Research Professor University of Tennessee Dr. Thomas S. Kress, Retired Oak Ridge National Laboratory Dr. Don W. Miller, Professor The Ohio State University ) Dr. William J. Shack Argonne National laboratory Dr. Robert E. Uhrig, Professor University of Tennessee Dr. Graham B. Wallis Dartmouth College l i Vii l l
l I l l l TABLE OF CONTENTS l l l East ABSTRACT ............................................ iii q PREFACE .............................................. v MEMBERSHIP .......................................... vii Potential Revision to 10 CFR 50.65(a)(3) of the Maintenance Rule to Require Licensees to Perform Safety Assessments, February 12,1998 ......................................... 1 Proposed Generic Letter, " Guidance on the Storage, Preservation, l and Safekeeping of Quality Assurance Records in Electronic
. Media," February 12,1998 ...................... ............ 3 Final Rulemaking, " Physical Protection for Spent Nuclear Fuel and High-Level Radioactive Waste" (10 CFR Parts 60, 72, 73, 74, and 75),
February 12 , 199 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l Interim Letter on the Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design, February 19,1998 ......................................... 7 ACRS Report to Congress on NRC's Safety Research Program,
- February 24,1998 ......................................... 17
- ACRS Comments on Draft Paper on Risk-Informed, Performance-Based 1 l Regulation, March 11,1998 .................................. 27 1
ix
F TABLE OF CONTENTS l ! P_aat i Proposed Final Standard Review Plan Sections and Regulatory l Guides for Risk-Informed, Performance-Based Regulation for Inservice Testing, Graded Quality Assurance, and Technical l Specifications, March 12,1998 ................................. 29 I Risk-Ranking Approach for Motor-Operated Valves, March 12,1998 ........ 33 Proposed Improvements to the Senior Management Meeting Process, March 13,1998..................................... 37 SECY-98-001, Mechanism for Addressing Generic Safety Issues, March 16, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 The Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design - Interim Letter 2, , April 9, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 Plans to Increase Performance-Based Approaches in Regulatory Activities, April 9, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 Proposed Rulemaking Changes to Paragraph (h) of 10 CFR 50.55a,
" Codes and Standards," April 9,1998 ............................ 53 Draft Commission Paper, " Status of the Environmental Qualification Task Action Plan," April 9, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 Proposed Final Generic Letter, " Year 2000 Readiness of Computer Systems at Nuclear Power Plants," April 9,1998 . . . . . . . . . . . . . . . . . . . . . 57 Draft Regulatory Guide DG-1078, " Standard Format and Content of License Termination Plans for Nuclear Power Reactors," April 9,1998 . . . . . 59 Proposed Final Amendment to 10 CFR Part 55, " Initial Licensed Operator Examination Requirements," May 8,1998 ................... 61 l
l l X l l l J
p I l l TABLE OF CONTENTS East Proposed Generic Ietter, " Modification of the NRC Staff's Recommendations for the Post-Accident Sampling System," May 8,1998 . . . . . . 63 Elevation of CDP to a Fundamental Safety Goal and Possible Revision of the Commission's Safety Goal Policy Statement, May 11,1998 .......... 65 Proposed Human Performance Plan and Proposed Final Standard Review Plan Section and Regulatory Guide for Risk-Informed Inservice Inspection of Piping, May 12,1998........................ 71 NRC Participation in the CABRI Reactor Fuels Research Program, ! June 9, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 Safety Evaluation Report on Electric Power Research Institute Topical Report, TR-107330, Final Report, " Generic Requirements l Specification for Qualifying a Commercially Available PLC for Safety-Related Applications in Nuclear Power Plants," June 9,1998 . . . . . . . . . 75 Draft Advance Notice of Proposed Rulemaking Regarding Amending 10 CFR 50.72, "Immediate Notification Requirements for Operating I Nuclear Power Reactors," and 10 CFR 50.73, " Licensee Event Report System," June 9,1998 ...................................... 77 Proposed Final Draft of the NRC's Human Performance Plan, i June 12, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 I Proposed Final Standard Review Plan Section 3.9.8 and Regulatory Guide 1.178 for Risk-Informed Inservice Inspection of Piping, June 12, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 NRC Reactor Fuels Research Program, June 15,1998 .................. 85 The Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design - Interim Letter 3, June 15, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 xi I i i
TABLE OF CONTENTS East Review of SECY-98-076, " Core Research Capabilities," June 16, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97 Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program, A Report to the U. S. Nuclear Regulatory Commission, NUREG-1635, Volume 1, June 1998 (Included by reference only) Proposed Revisions to 10 CFR 50.59 (Changes, Tests and Experiments), July 16,1998................................... 99 Draft Supplement 1 to NUREG 1552, " Fire Barrier Penetration Seals in Nuclear Power Plants," July 20,1998 ....................... 111 Proposed Final Safety Evaluation of the BWR vessel and Internals Project, BWR Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05) Report, July 21,1998 ................. 115 Draft Regulatory Guide DG-4005, " Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses," July 22,1998 .......................... 117 Report on the Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Passive Plant Design, July 23,1998 ................................... 119 Proposed Revision to 10 CFR 50.65(a)(3) of the Maintenance Rule to Require Licensees to Perform Safety Assessments, July 22,1998 ......... 131 General Electric Nuclear Energy Extended Power Uprate Program and Monticello Nuclear Generating Plant Power Level Increase ! 133 I Request, July 24,1998 ...................................... Proposed Final Rulemaking: Changes to Paragraph (h) of 10 CFR Part 50.55a, " Codes and Standards," September 9,1998 . . . . . . . . . . . . . . . . . 139 xii i l
I I l TABLE OF CONTENTS Ease l Emergency Core Cooling System Strainer Blockage, September 14,1998...... 141 Application for Power Level Increase for Edwin I. Hatch Nuclear Power Plant Units 1 and 2, September 15, 1998 . . . . . . . . . . . . . . . . . . . . . . 145 i Impact of Probabilistic Risk Assessment Results and Insights on ' the Regulatory System, September 30,1998......................... 149 Proposed Final Amendment to 10 CFR Part 55, " Initial Licensed . I Operator Examination Requirements," October 6,1998 . . . . . . . . . . . . . . . . . 155 i Proposed Rule " Changes to Requirements for Environmental J Review for Renewal of Nuclear Power Plant Operating Licenses (10 CFR Part 51)," October 6,1998 ............................. 157 i Proposed Final Revisions to Regulatory Guide (RG) 1.84 (Design and Fabrication Code Case Acceptability), RG 1.85 (Materials Code Case Acceptability), and RG 1.147 (Inservice Inspection Code Case l Acceptability), October 7,1998 ................................ 159 i Proposed Final Regulatory Guide DG-1029, Revision 1.7,
" Guidelines for Evaluating Electromagnetic and Radio-Frequency i Interference in Safety-Related Instrumentation and Control Systems," i October 8, 1998 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 161 Risk-Informed Pilot Application for Hydrogen Monitoring at Arkansas Nuclear One, Units 1 and 2, October 14, 1998 . . . . . . . . . . . . . . . . . . . . . . . 163 Proposed Priority Rankings of Generic Safety Issues: Tenth Group, October 16,1998........................................... 165 The Nuclear Energy Institute's Petition for Rulemaking to Amend Paragraph (a) of 10 CFR 50.54, Conditions of Licenses, October 20,1998 . . . . . 179 xiii l
l
TABLE OF CONTENTS P3Et Proposed Amendments to 10 CFR 50.47: Granting of Petitions for i Rulemaking (PRM 50-63 and 50-63A) Relating to a Reevaluation of I I Policy on the Use of Potassium Iodide (KI) After a Severe Accident at a Nuclear Power Plant, November 12,1998......................... 181 Proposed Final Generic Letter, " Laboratory Testing of Nuclear-Grade l 1 Activated Charcoal," November 12,1998 .......................... 183 1 SECY-98-253, " Applicability of Plant-Specific Backfit Requirements to Plants Undergoing Decommissioning," November 13,1998 ............. 185 I Draft Commission Paper Concerning Initiation of Rulemaking - Revision of 10 CFR 55.31(a)(5) and 55.45(b) Regarding the Use ; of Simulators in Operator Licensing, November 13,1998 ................ 187 Proposed Inspection Procedure 35XXX, " Graded Quality Assurance," November 13,1998 ........................................ 189 Proposed Revision to the Enforcement Policy, November 17,1998 .......... 191 Proposed Rule on Use of Alternative Source Term at Operating Reactors, November 19,1998 ........................................ 195 Safety Evaluation Report Related to Westinghouse Owners Group Application of Risk-Informed Methods to Inservice Inspection of Piping, Topical Report (WCAP-14572, Revision 1), November 20,1998 . . . . . . 199 Reprioritization and Proposed Resolution of Generic Safety Issue-171,
" Engineered Safety Features Failure from Loss-of-Offsite-Power Subsequent to a Loss-of-Coolant Accident," November 23,1998 ........... 205 Options for Incorporating Risk Insights Into the 10 CFR 50.59 Process, December 11,1998 ........................................ 209 xiv
l TABLE OF CONTENTS i l East Proposed Commission Paper Concerning Options for Risk-Informed Revisions to 10 CFR Part 50 " Domestic Licensing of Production and Utilization Facilities," December 14,1998.......................... 211 Rulemaking Plan - Protection Against Discrete Radioactive Particle (DRP) Exposures (10 CFR Part 20), December 14,1998 ................ 215 Proposed Improvements to the NRC Inspection and Assessment Programs - Interim Report, December 16,1998 ...................... 217 xv
%, UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION '
V, ADVISORY COMMITTEE ON REACTOR SAFECUARDS WASHINGTON, D. C. 20666 February 12,1998 MEMORANDUM TO: L Joseph Callan Executive Director' C - c-- FROM: John T. Larkins, Director ACRS/ACNW
SUBJECT:
POTENTIAL REVISION TO 10 CFR 50.65(a)(3) OF THE MAINTENANCE RULE TO REQUIRE LICENSEES TO PERFORM SAFETY ASSESSMENTS Dunng the 448th meetmg of the Advisory Committee on Reactor Safeguards,
^ February 5-7,1998, the Committee considered the potential revision to 10 CFR 50.65(a)(3) of the Maintenance Rule. The Committee decided to review this matter following reconciliation of public comments.
References-
- 1. SECY-97-173, dated August 1,1997, for the Commissioners, from L. Joseph Callan, Executive Director for Operations,
Subject:
Potential Revision to 10 CFR 50.65(a)(3) of the Maintenance Rule to Require Licensees to Perform Safety Assessments
- 2. Staff Requirements Memorandum dated December 17,1997, from John C. Hoyle, Secretary, to L. Joseph Callan, Executive Director for Operations, and Karen D. Cyr, GeneralCounsel,
Subject:
Staff Requirements: SECY-97-173 - Potential Revision to 10 CFR 50.65(a)(3) of the Maintenance Rule to Require Licensees to Perform Safety Assessments cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO M. Knapp, RES S. Collins, NRR S. Black, NRR J. L. Funches, CFO H. T. Bell, OlG 1
# 'o,, UNITED STATES
/ o NUCLEAR REGULATORY COMMISSION d' ,I ADVISORY COMMITTEE ON REACTOR SA,FEGUARDS o
WASMNGTON. D. C. 20666 February 12,1998 MEMORANDUM TO: L Joseph Callan Execubve Dir orf FROM: John T. Larkins, e ' rector ACRS/ACNW
SUBJECT:
PROPOSED GENERIC LETTER," GUIDANCE ON THE STORAGE, PRESERVATION, AND SAFEKEEPING OF QUALITY ASSURANCE RECORDS IN ELECTRONIC MEDIA" During the 448th meeting of the Advisory Committee on Reactor Safeguards, February 5-7,1998, the Committee decided not to review the proposed subject generic letter. Reference-Memorandum dated January 9,1998, from Frank J. Miraglia, NRR, to Thomas T. Martin, CRGR,
Subject:
Request for Review and Endorsement of the Proposed Generic Letter Titled,
" Guidance on the Storage, Preservation, and Safekeeping of Quality Assurance Records in Electronic Media."
cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR S. Black, NRR T. Martin, AEOD M. Knapp, RES 3
p easp
'o UNITED STATES g
! o NUCLEAR REGULATORY COMMISSION I I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS i / WASHINGTON. D. C. 20555 e ss,e* February 12,1996 MEMORANDUM TO: L Joseph Callan Executive Director f FROM: JohnT. Larkins e Director ACRS/ACNW
SUBJECT:
FINAL RULEMAKING, " PHYSICAL PROTECTION FOR SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE"(10 CFR PARTS 60,72,73,74, AND 75) During the 448th meeting of the Advisory Committee on Reactor Safeguards, February 5-7,1998, the Committee decided not to review the subject final rule.
Reference:
Memorandum dated January 9,1998, from Elizabeth Q. Ten Eyck, Director, Division of Fuel Cycle Safety and Safeguards, NMSS, to Addressees,
Subject:
Final Rulemaking, Physical Protection for Spent Nuclear Fuel and High-Level Radioactive Waste cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO M. Knapp, RES E. Ten Eyck, NMSS P. Dwyer, NMSS J. L Funches, CFO H. T. Bell, OlG 5
# o,, UNITED STATES 8 o NUCLEAR RECULATORY COMMISSION y I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS S mamworow, p. c.nassa February 19, 1998 Mr. L. Joseph Callan Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Mr. Callan:
SUBJECT:
INTERIM LETTER ON THE SAFETY ASPECTS OF THE WESTINGHOUSE ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE AP600 PLANT DESIGN During the 448th meeting of the Advisory Comittee on Reactor Safeguards, February 5-7, 1998, we reviewed the AP600 test and analysis program and various chapters of the AP600 Standard . Safety Analysis Report. Our Subcommittees on Advanced Reactor Designs and on Thermal Hydraulic and Severe Accident Phenomena have reviewed these matters previously, as listed in Attachment 1. During these reviews, we had the benefit of discussions with representatives of the NRC staff and the Westinghouse Electric Company (Westinghouse) and of the documents referenced. TEST AND ANALYSIS PROGRAM The central goals of the Westinghouse Test and Analysis Program (TAP) are to confirm the design basis for the nuclear power plant components and systems unique to the AP600 design and to provide test data to support validation of relevant plant system codes. Westinghouse has concluded its testing programs, and its current focus is on the verification of the pertinent analytical tools. Our Thermal Hydraulic and Severe Accident Phenomena Subcommittee began its review of the Westinghouse TAP in December 1991 and several meetings of the Subcommittee have been held in the interim. The Subcomittee last met to review the status of the key elements of the Westinghouse TAP on December 9-12, 1997. Over the course of these reviews, the Thermal Hydraulic and Severe Accident Phenomena Subcommittee has raised a number of issues that have been documented only in Subcomittee minutes and transcripts, Subcommittee Chairman's reports, and ACRS consultants
- reports. In the interest of documenting these issues in a single report. a listing is provided below. We recomend that Westinghouse 7
Mr. L. Joseph Callan provide responses on these issues to the NRC staff for review. Those issues that we consider to be of higher priority are marked with an asterisk. Reactor Coolant System Issues:
- The basis for not including the momentum fluxes in the NOTRUMP code, particularly during the blowdown phases of the accident analyses ,
- Explanation of the applicability of Equation 3-63 of Reference 3 to the critical flow of a single component two-phase fluid
- Validation basis for the drift-flux modeling of horizontal flow
- Explanation of why the blowdown flows out of the automatic depressurization system (ADS) valves 1. 2. and 3 and out of the break itself are not predicted well by the NOTRUMP code and what will be done to assure a conservative prediction of AP600 behavior (we are particularly concerned about using modeling deficiencies as compensating effects)
- A more complete demonstration that the proposed penalty on fluid level in the in-containment refueling water storage tank (IRWST) provides sufficient conservatism to offset the uncertainties in the calculated pressurizer level holdup and resulting minimum core level
- The basis for validation for the liquid entrainment model used for the ADS-4 line
- Justification for the absence of (or completion of) a " multi-loop" scaling analysis during the IRWST cooling phase when the vessel inventory approaches a minimum
. Description of the pressurizer flooding model and its validation basis (treatment of the surge line from the hot leg to the pressurizer)
. Explanation of how upstream flow effects were treated in reducing the data in the ADS separate effects tests and in the NOTRUMP code
. The basis for the inconsistencies between the NOTRUMP code noding used for the integral system test configurations and that used for the AP600 plant model 8
Mr. L Joseph Callan Containment Issues:
- Justification for the use of an incorrect expression in the rate-of-pressure-change equation (Equation 34 in Reference 6)
- Justification for the inappropriate cancellation of the partial derivative of internal energy at constant pressure by the partial derivative of internal energy at constant volume to arrive at Equation 34 of Reference 6
- Re-evaluation of the derivation and quantification of the scaling pi groups resulting from a correction of Equation 34 of Reference 6
- Justification for using the WG0THIC lumped parameter model well-mixed assumption for calculating the AP600 containment behavior e Justification for the use of steady-state testing in the Passive Containment System Large Scale Test facility to validate transient heat transfer correlations in the WG0THIC code o Justification for the normalization of the rate-of-pressure-change term in Equation 34 in Reference 6 o Technical basis for the treatment of the cooled containment boundary laminar sublayer in the WGOTHIC code o Validation basis for assuming a low elevation for the main steam line break o Justification that the calculated peak containment pressure has appropriate margin in view of the observation that all three of the containment cooling system mechanisms (i.e., the passive cooling water system, heat transfer to the containment shell. and heat transfer to the internal structures) are required to turn the pressure over just as it reaches the design value o Quantification of the impact of incorrect (with respect to AP600) relative magnitudes of energy and mass addition and energy removal during the Large Scale Tests on the usefulness of the data for WGOTHIC code validation for use on AP600 9
Mr. L Joseph Callan In addition to the above, we are disturbed by the poor status of documentation related to information needed to certify the AP600 design. We believe that any certification should be contingent upon documentation of sufficient quality to provide a traceable and well-archived licensing basis. SAFETY ANALYSIS REPORT Our Advanced Reactor Designs Subcommittee began its review of the AP600 design in January 1995. Since then, we have issued two reports to the Commission: one report concerned policy and key technical issues, and the second supported the requirement for a containment spray system. We have reviewed the following Standard Safety Analysis Report chapters and have no comments at this time:
. Chapter 1 - Introduction . Chapter 4 - Reactor . Chapter 5 - Reactor Coolant and Connected Systems . Chapter 7 - Instrumentation and Controls . Chapter 8 - Electrical Power . Chapter 11 - Radioactive Waste Management . Chapter 13 - Plant Operations (excluding security) l . Chapter 18 - Human Factors Engineering
SUMMARY
. We have identified a number of issues associated with the Westinghouse Test and Analysis Program that should be resolved during the staff review. Our assessment of the adequacy of the Standard Safety Analysis Report chapters discussed to date is incomplete. Completion of our review is contingent on the timely receipt of draft Final Safety Evaluation Report chapters.
Sincerely. Robert L. Seale Chairman 10
Mr. L. Joseph Callan
References:
- 1. Westinghouse Electric Corporation. "AP600 Standard Safety Analysis Report," updated through Revision 16 dated September 2, 1997.
- 2. Letter dated January 16, 1998. from William Huffman. NRC, to Nicholas Liparulo, Westinghouse Electric Corporation.
Subject:
Open Items Associated with the AP600 Safety Evaluation Report on the AP600 Containment Design and Accident Analyses.
- 3. Westinghouse Electric Corporation, WCAP-14727, Revision 1. "AP600 Scaling and PIRT Closure Report." July 1997 (Proprietary).
- 4. Westinghouse Electric Corporation, WCAP-10079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985 (Proprietary).
- 5. Westinghouse- Electric Corporation. WCAP-10054-P-A. " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code." August 1985 (Proprietary).
- 6. Westinghouse Electric. Corporation. WCAP-14845, Revision 2: " Scaling Analysis for AP600 Containment Pressure During Design Basis Accidents "
June 1997 (Proprietary).
- 7. Westinghouse Electric Corporation, WCAP-14407 Revision 1. "WG0THIC Application to AP600 " July 1997 (Proprietary).
- 8. Westinghouse Electric Corporation, WCAP-14326. Revision 1. " Experimental Basis for the AP600 Containment Vessel Heat and Mass Transfer 4' Correlations." May 1997 (Proprietary).
- 9. Westinghouse Electric Corporation. WCAP-14807 Revision 2. "NOTRUMP Final Validation Report for AP600," June 1997 (Proprietary).
- 10. Westinghouse Electric Corporation, WCAP-14967. Revision 0. " Assessment of Effects of WG0THIC Solver Upgrade From Version 1.2 to 4.1 " September 1997 (Proprietary).
- 11. Westinghouse Electric Corporation WCAP-14135. " Final Data Report for PCS Large-Scale Tests. Phase 2 and Phase 3." July 1994 (Proprietary).
Attachment:
- 1. Chronology of the ACRS Review of the Westinghouse Application for AP600 Standard Design Certification
ATTACHMENT 1 CHRONOLOGY OF THE ACRS REVIEW 0F THE WESTINGHOUSE APPLICATION FOR AP600 STANDARD DESIGN CERTIFICATION SUBCOPNITTEE MIE SUBJECT Thermal Hydraulic Phenomena Review of Proposed Comission Paper on Need for 12/17/91 Full-Height. Full-Pressure Integral System Testing of AP600 Design Thermal Hydraulic Phenomena Continue Review of Integral System Testing 3/3/92 Requirements for AP600 Passive Plant Design Thermal Hydraulic Phenomena Continue Review of Integral System Testing 6/23-24/92 Requirements for AP600 Passive Plant Design Thermal Hydraulic Phenomena Continue Review of Westinghouse Test and 7/22-23/93 Analysis Program for AP600 Design Thermal Hydraulic Phenomena Continue Review of Westinghouse Test and l 9/21/93 Analysis Program - Oregon State University Test Facility Thermal Hydraulic Phenomena. Continue Review of Westinghouse Test and 3/15-16/94 Analysis Program - Core Makeup Tank Test Facility l W Standard Plants Designs Overview Chap. 1: Introduction and 1/11/95 General Description of Plant Thermal Hydraulic Phenomena Review of COBRA / TRAC codes for W AP600 2/15-16/95 Thermal Hydraulic Phenomena Review test and analysis programs for AP600 3/29-30/95 Passive Containment Cooling System W Standard Plant Designs Review Staff Commission Paper on Status of Ten 5/31/95 Key Technical and Policy Issues 1 [ 13
Thermal Hydraulic Phenomena Staff review of Qualification Documentation for 7/26-27/95 the W COBRA / TRAC code Thermal Hydraulic Phenomena Staff review of Qualification Documentation for 1/18-19/96 the W COBRA / TRAC code W Standard Plant Designs SECY-96-128, " Policy and Key Technical Issues 7/19/96 Pertaining to the AP600 Design" 433rd ACRS Meeting SECY-96-128. " Policy and Key Technical Issues" 8/8/96 ACRS Report Issued 8/15/96 W Standard Plant Designs Chap. 4: Reactor 12/4/96 Chap. 5: Reactor Coolant System and Connected Systems Chap. 9: Auxiliary Systems Chap. 11: Radioactive Waste Management Thernal Hydraulic Phenomena Scaling and PIRT Closure Report 12/18-19/96 Thermal Hydraulic Phenomena Test and Analysis Program: Long-Term Core 3/28/97 Cooling With W COBRA / TRAC Code 442nd ACRS Meeting AP600 Containment Spray Design 6/13/97 ACRS Report issued 6/17/97 Thermal Hydraulic Phenomena NOTRUMP Small-Break LOCA Code 7/29-30/97 , Thermal Hydraulic Phenomena Passive Containment System Test and Analysis 9/29-30/97 Program Thermal Hydraulic Phenomena PIRT: Scaling of RCS: NOTRUMP Small-Break LOCA 12/9-10/97 Code Thermal Hydraulic Phenomena WGOTHIC Containment System Code 12/11-12/97 Advanced Reactor Designs Chap. 1: Introduction 2/4/98 Chap. 4: Reactor 14
Chap. 5: Reactor Coolant System and Connected Systens Chap. 7: Instrumentation and Controls Chap. 8: Electrical Power Chap. 11: Radioactive Waste Management Chap. 13: Conduct of Operations Chap. 18: Human Factors Engineering 448th ACRS Maeting TAP and SSAR Chapters 1. 4. 5. 7, 8. 11. 13. and 18 Interim Letter issued February 19, 1998 15
# #g UNITED STATES 8 g NUCLEAR REGULATORY COMMISSION
{ r ADVISORY COMMITTEE ON REACTOR SAFEGUARDS $ WASHINGTON, D. C. 20555 February 24,1998 The Honorable Albert Gore, Jr. President of the United States Senate Washington, DC 20510 Dear Mr. President I am pleased to transmit to the Congress the 1997 report of the Advisory Committee on Reactor Safeguards on the U.S. Nuclear Regulatory Commission's Safety Research Program. This report is required by Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209. This report concludes that severe budget reductions are causing substantial deterioration of the intemationally respected capability of the U.S. Nuclear Regulatory Commission to conduct a forward-looking, effective safety research program. Sincerely, R. L Seale Chairman
Enclosure:
Nuclear Safety Research, A Report to the U.S. House of Representatives and the U.S. Senate, by the Advisory Committee on Reactor Safeguards of the U.S. Nuclear Regulatory Commission, dated February 1998 17
*g UNITED STATES
/ , NUCLEAR REGULATORY COMMISSION y -
ADVISORY COMMITTEE ON REACTOR SAFE 3UARDS q WASHINGTON, D. C. 20585
***** February 24,1998 i
The Honorable Newt Gingrich Speaker of the United States House of Representatives Washington, DC 20515
Dear Wir. Speaker:
I am pleased to transmit to the Congress the 19g7 report of the Advisory Committee on Reactor Safeguards on the U.S. Nuclear Regulatory Commission's Safety Research Program. This report is required by Section 29 of the Atomic Energy Act of 1954, as amended by Sechon 5 of Public Law 95-209. This report concludes that severe budget reductions are causing substantial deterioration of the intemationally respected capability of the U.S. Nuclear Regulatory Commission to conduct a forward-looking, effective safety research program. Sincerely, R. L Seale Chairman
Enclosure:
Nuclear Safety Research, A Report to the U.S. House of Representatives and the U.S. Senate, by the Advisory Committee on Reactor Safeguards of the U.S. Nuclear Regulatory Commission, dated February 1998 18
NUCLEAR SAFETY RESEARCH A Report to the U.S. House of Representatives and the U.S. Senate by the Advisory Committee on Reactor Safeguards ofthe U.S. Nuclear Regulatory Commission I February 1998 19
The Atomic Energy Act of 1954, as amended by Seebon 5 of Public Law 95-209, requires that the Advisory Committee on Reactor Safeguards of the U.S. Nuclear Regulatory Commission report annually to Congress on the status of nuclear reactor safety research. This is the 1997 report of the Advisory Committee on Reactor Safeguards. This report concludes that severe budget reductions are causing substantial deterioration of the intomationally respected capability of the U.S. Nuclear Regulatory Commission to conduct a forward-looking, effechve safety research program. As we described in the report of 1996, this detenorabon is occurring at a time when the U.S. nuclear power industry is undergoing substantial changes in response to economic deregulation made possible by the Energy Policy Act of 1992. These changes may have safety implications that must be addressed by the Commission. Research is needed to ensure that the agency effectively addresses these changes. The deteriorabon in recearch capabilities is also inhibiting the ability of the Commission to continue the evolution of riuclear reactor regulation to a risk-informed, performance-based structure. Finally, the Cornmission's core capability in nuclear waste research has been dramatically reduced. Further reduchons could inhibit the Commission staff's effectiveness and timeliness in conducting reviews of the nuclear waste repository program and cause delays and additional expenditure of National resources.
Background
The use of nuclear energy to provide electricity to the civilian population was pioneered in the United States. This technology has now spread among the developed nations of the world and all indications are that it will aiso be adopted by developing nations in the future. Today, the majority of the 450 operating nuclear power plants and plants under construction throughout the world are based on U.S. technology, it was, of course, well recognized in the initial applications of nuclear energy for civilian purposes that the health and safety of the public must be adequately protected. Because there was at the time so little experience with such a new technology, very conservative, prescriptive regulations emphasizing a defense-in-depth approach to safety were established to control the civilian use of 1 20
nuclear power and the management of nuclear waste. Overly conservative regulations that do not have safety significance serve only to inhibit the fruitful applicaten of the technology. Congress recognized, however, that even the most stringent regulations might not anticipate all the safety issues of a new iWJi,0l0gy. Congress, therefore, encouraged safety research to further develop and refine the regulabon of nuclear power. Recently, Congress has encouraged all regulatory agencies, includmg the Nuclear Regulatory Commission, to assess and refine regulatory schons j to ensure that the costs and burdens imposed by regulatory schons are commensurate with the derived isocietal benefit. Sam the earty days, nuclear power has become an essential, reliable contributor to the Nation's energy suppies. Today, nuclear energy provides about 20 percent of the overall electrical energy in the country, and it does so with very low emissions of particulate and gaseous pollutants. There , are regions of the country where nuclear power is the dominant source of electncal energy. In some countnes, nuclear power is an even more important source of electrical energy. Along with its role in the development and dissemination of this technology, the U.S. has become the world leader in nuclear safety. This leadership is due in no small part to the thorough safety research that the U.S. Nuclear Regulatory Commission has been able to perform in the past. This safety I technology contnbutes significantly to the acceptance and purchase of U.S. nuclear technology by other nations. Well researched, well maintained standards and regulations for the nuclear fuel cycle, such as those developed by the Commission, reduce the potential for the proliferation of nuclear weapons matenals as the worldwide use of nuclear power expands. The Situation Today While use of nuclear energy in the United States is not growing, the U.S. nuclear industry is by no means static. The industry is, in fact, undergoing substantial change. Changes due to modem technical developments such as the
- digital revolution" in the instrumentabon and control of nuclear reactors are to be expected and will improve both safety and efficiency if properly impismented.
The changes that occur as nuclear power plants age must be addressed to ensure continued safety and reliability. Of more importance, and a definite source of greater uncertainty, is the change in the nuclear industry caused by economic deregulation. The pressures of increased competition will produce changes that could well have safety implications. Certainly, steps taken , 21
by the industry to reduce manpower and to enhance the productivity of the remaining personnel need to be scruhnized closely and researched for safety significance. Similar comments can be made about steps being taken by the nuclear industry to extend the lifetime of nuclear fuel and to dh.;ff the suppliers of nuclear fuel for indmdual plants. ! l The nuclear industry also faces the challenge in the future of a growing volume of spent nuclear fuel. Repositories for the disposal of nuclear wastes are prerequisites for the sustained use of nuclear power. Radioactive disposal facilities will also be crucial for continued use of nuclear materials in medicine, other industries, and scientific research. l i if regulabon is not to stifle economic and technical improvements in the U.S. nuclear industry, tie ] U.S. Nuclear Regulatory Commission must be in a position to modemize its safety regulations. ) Modomization will also be essential if the United States is to maintain its world leadership as a supplier of both nuclear technology and nuclear safety technology. There are now about 3,000 reactor-years of operatonal experience in the commercial use of nuclear power. The Commissen has fostered through research the development and refinement of systematic methods to collect operational safety data, to assess these data, and to combine the data sets into integrated I evaluations of the safety of nuclear power plants. On the basis of the data and analyses, the Commission is now undertaking an important evolution of its regulations to a risk-informed, performance-based structure. The Commission is the leader, in fact, among this country's regulatory agences and within the world's nuclear regulatory agencies at rational regulation that focuses efforts on topics of the greatest safety significance and assures that regulations are commensurate with the derived societal safety benefits. Over the last year, the U.S. Nuclear Regulatory Commission has initiated changes to its regulatons and practices that encourage risk-informed considerations to be taken in the development of technical specifications for nuclear power plants, in-service inspection and testing of safety systems, and graded quality assurance programs for safety systems. Pilot applications of these efforts to improve regulations are being conducted and the results are now being assessed. These moves toward risk-informed regulation are expected to improve safety and regulatory officency. They are also expected to reduce costs to the nuclear industry and to the American public. For example, the recent move to performance-based containment leak rate testing is 3-22
cxpected to produce cost savings approaching a billion dollars over the projected lifetimes of cxistmo plants. Innovatens being made today by the U.S. Nuclear Regulatory Commission in its regulabons are made possible by the research that has been done in the past. The Commission has, in fact, a good record regarding the prudent identificabon of important safety research issues and the effective conduct of research. During the last year, for, example, past results from the research program have enabled the Commission to assess industry arguments concoming required inspectons of reactor vessel welds. Potential problems identified by the research program have led to requirements for additonal attenton to the qualificaten of motor-operated valves in existing nuclear power plants. Past research has also made possible the certificahon of two new nuclear power plant designs: the General Electnc Advanced Boiling Water Reactor and the ABS-CE System 80+ pressurned water reactor. The research program is contributing to the evaluation of the advanced light water reactor design now being proposed by Westinghouse for certification by the Commiss'en. The Crisis in Nuclear Safety Research Despite the substantial changes the nuclear industry is undergoing, the budget available for the conduct of regulatory activibes by the Nuclear Regulatory Commissen is decreasing. In the face of deciming resources, prionty, of course, must be given to operational actrvibes such as effective monitoring and inspection of licensees and the disposition of current licensing actions. Recent, well-publicaed events at partcular nuclear facilities have underscored the priority that needs to be given to such continuing vigilance. Consequently, many of the longer term benefits that could come from research have had to be deferred. The resources available for research have decreased disproportionately in the last several years. The research program has sustained reduchons of 23 percent in 1996,19 percent in 1997, and 16 percent in 1998. The declining i resources available for needed research are having impacts now. Examples include: o A program to monitor industry research and to anticipate initiatives that may require revisions of regulations in the future has not been undertaken. The Nuclear Regulatory Commission is being forced into a position where it must wait and react to indestry ) 23
proposals and thereby delay implementation of innovabons even if these initiatives improve safety. Delays have already been encountered in the implementation of revised accdont source terms and new dosimetry methods because the Commisskm cannot afford to complete needed research. Delays caused by deferred research on risk-informed pilot progects have detressed the nuclear mdustry whose hopes for improved regulabons in the near future have begun to dwindle. o Research needed to evaluate the potenbal for safety-sigrvficant human errors, especially as the nuclear industry " downsizes" staff in response to economic deregulation, remains in the planning stages despde contmung evdence from plant opershons that human errors are important contnbutors to off-normal events at nuclear power plants. o The technology has not been developed to extend systematic evaluabons of risk from normal power plant opershons to shutdown and low power operatons despite evdence that these are modes of opersbon that pose risk to the public comparable to that from power operabons. o Research needed to evaluate hcensee proposals to extend the lifetime of reactor fuel, whch will also reduce the societal burden of spent nuclear fuel, remains to be performed. I o Safety research that will be needed to regulate the use of mixed oxide fuels as a means for the disposal of the Nation's excess weapon grade plutonium has not been initistM. o The program to develop a technical understanding of public health and safety risks posed by severe reactor accdonts may have to be terminated prematurely. Research on the safety and risk sigrvficance of fires has been deferred. The ability of the Commission to leverage dwindling research resources by collaborabon in initiatives by other countries with more ambitious research programs may be jeopardized o Validation of industrial standards to use in place of Govemment-formulated regulations will be slowed. 24
o Key elements of a welkiesigned research program to assist in the licensing of a high-level nuclear ' waste repository are being adversely impacted by Congressional funding reduchons. Without the research results that reduce uncertainties, it may be necessary to add conservabsm, and thus raise costs for the design of the waste repositories to ensure adequate premi of the public health and safety. o Fifteen of the genene safety issues identified since the 1979 amendments to the Energy Reorganizabon Act of 1974 have still not been resolved. Deficiencies in the research program that the U.S. Nuclear Regulatory Commission can afford to maintain will affect the performance of line organizations responsible for ongoing regulatory activibes with bconsees. Even today, requests or " user needs" for research by line organizations are being withhold because it is known that the reduced research program cannot respond to such requests. Of concem now are limitabons developing in the ability of the research program to conduct systematic examinations of the effechveness of existing regulations and to identify additional areas for risk-informed, performance-based improvements. There are also concems cbout the availability of financial resources to sustain safety research on emerging digital bri-A-j:1 Without advanced safety research, applicahon of these superior technologies to the instrumentabon and control of nuclear power plants will be delayed, along with attendant improvements in safety and plant performance. Conclusions The U.S. Nuclear Regulatory Commission and the safe regulation of nuclear power plants have benefited from research done in the past. Reductions in the Commission budget heve forced serious cutbacks in the research program and deterioration of the research capability. The Commission still needs a research program. It certainly needs a viable program to be able to evaluate proposals independently and to assess safety arguments advanced by the industry. It needs a stronger research program to continue the evolution of its safety regulations. The Commission also needs a research program to meet new obligations it is undertaking. Notable among the new obligations is the implementation of safety regulations for a geologic repository for 25
spent nuclear fuel. The agency is also conducting a pilot program to assess the viability of undertaking the safety regulation of certain Department of Energy nuclear facilities. The Nuclear Regulatory Commission capacity for research is no longer commensurate with the agency's regulatory obligations. It will not be possible to maintain core competencies in all the areas that have historically proven to be of recumng imperw.cs in safety and regulatory actions by the egency. Modomizabon of regulabons wiH be delayed because research cannot be performed to ensure that appropriately high levels of safety are maintained. Responses to industrial initiabves taken as a result of compebtrve pressures wiH be slowed without a broader research program. Delay in the implementabon of cost compeutive innovabons may well force the nuclear industry to rebre more plants prematurely, and the Nabon wiu incur at the socetal costs such unnecessary retirements entail. The development of a high-level nuclear waste repository is facilitated by the availability of well-researched safety regulations and analytical tools for licensing. Uncertainties left when research cannot be done because of funding constraints may delay the development of the repository or force the addition of costly conservatism. In summary, there are benefits to the entire socoty that may be delayed or even lost as the research capability of the U.S. Nuclear Regulatory Commission deteriorates in response to declining financial resources. I 26
e ase UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION .$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS S WASHINGTON, D. C. 20065 March 11. 1998 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
ACRS COMMENTS ON DRAFT PAPER ON RISK-INFORMED. PERFORMANCE-BASED REGULATION During the 449th meeting of the Advisory Comittee on Reactor Safeguards. March 2-4. 1998, the Comittee considered your February 20. 1998 memorandum and attached draft paper intended to define and clarify several concepts related to risk-informed, performance-based regulation. Our suggested comments and recomendations are provided in the attached redlined version of the subject document. We appreciate the opportunity to provide comments on this paper. We believe agreement on definition of terms will be important to moving forward on this subject. Sincerely. R. L. Seale Chairman
Reference:
Memorandum dated February 20, 1998, from Shirley Ann Jackson. Chairman. NRC, to Robert L. Seale Chairman. ACRS.
Subject:
Discussion of Risk-Informed. Performance-Based Regulation
Attachment:
As Stated [Not included] cc: Commissioner Dicus Commissioner Diaz Commissioner McGaffigan 27
/ 'o UNITED STATES ! n NUCLEAR REGULATORY COMMISSION j- i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o WASHINGTON, D. C. 20666 March 12,1998 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
PRDPOSED FINAL STANDARD REVIEW PLAN SECTIONS AND REGULATORY GUIDES FOR RISK-INFORMED. PERFORMANCE-BASED REGULATION FOR INSERVICE TESTING. GRADED QUALITY ASSURANCE. AND TECHNICAL SPECIFICATIONS During the 449th meeting of the Advisory Committee on Reactor Safeguards March 2-4, 1998, we met with representatives of the NRC staff to review proposed final ; Standard Review Plan (SRP) sections and regulatory guides for risk-informed, l performance-based regulation including individual applications for inservice testing, graded quality assurance, and technical specifications. We discussed the staff's reconciliation of public comments on the subject documents. Our Subcommittee on Reliabihty and Probabilistic Risk Assessment met with the staff and industry representatives on February 19, 1998, to discuss these matters. We also had the benefit of the documents referenced. Conclusions and Recommendations
- 1. We recommend that Regulatory Guides 1.175 (Inservice Testing). 1.176 (Graded Quality Assurance), and 1.177 (Technical Specifications) and associated SRP sections be approved and issued for use.
- 2. We do not believe that Regulatory Guide 1.176 takes full advantage of the information that probabilistic risk assessment (PRA) provides. We recognize, however, that the lack of a model for assessing the l quantitative impact of quality assurance requirements on PRA parameters makes this a particularly difficult document to write.
- 3. We recommend that the Office of Nuclear Regulatory Research consider a research project to assess the impact of quality assurance requirements on PRA parameters.
29
h 2
- 4. We recommend that the staff prepare a plan for improvements to Regulatory Guide 1.176 after experience with its application and related studies and l brief the Comittee sometime in the next two years.
As stated in our previous reports, we believe that the next major step in the process will be the use of these documents in practice. We urge the staff to l move expeditiously to reach closure on the pilot risk-informed requests for l changes to the current licensing basis that are currently under review. We were l pleased to hear a presentation from the Nuclear Energy Institute on the new risk-l informed initiative that it is sponsoring. We plan to follow developments in j these activities with great interest. Sincerely, R. L. Seale Chairman
References:
- 1. U.S. Nuclear Regulatory Comission, proposed final SRP Section 3.9.7.
" Risk-Informed Inservice Testing." draft dated March 2. 1998 (Predecisional).
- 2. U.S. Nuclear Regulatory Comission, proposed final Regulatory Guide 1.175, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Inservice Testing.~ draft dated March 2, 1998. (Predecisional)
- 3. U.S. Nuclear Regulatory Comission, proposed final SRP Chapter 16.1,
" Risk-Informed Decisionmaking: Technical Specifications," draft dated March 2, 1998 (Predecisional).
l
- 4. U.S. Nuclear Regulatory Comission, proposed final Regulatory Guide l'.176, l "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Graded Quality Assurance " draft dated March 2, 1998 (Predecisional).
- 5. U.S. Nuclear Regulatory Commission. proposed final Regulatory Guide 1.177.
"An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications." draft dated March 2,1998 (Predecisional).
- 6. Report dated March 17, 1997, from R. L. S3 ale Chairman ACRS. to Shirley Ann Jackson, Chairman NRC.
Subject:
Propo:,cd Standard Review Plan Sections and Regulatory Guides for Risk-Informed. Performance-Based Regulation.
- 7. Report dated December 11, 1997, from R. L. Seale, Chairman. ACRS, to Shirley Ann Jackson, Chairman. NRC
Subject:
Proposed Final Regulatory Guide 1.174 and Standard Review Plan Chapter 19 for Risk-Informed. Performance-Based Regulation. I 30
3
- 8. Memorandum dated October 30. 1997. from John C. Hoyle. Secretary of the Commission, to L. Joseph Callan. Executive Director for Operations. NRC.
Subject:
Staff Requirements Memorandum - S'!CY-97-229. " Graded Quality Assurance /Probabilistic Risk Assessment Implementation Plan for the South Texas Project Electric Generating Station."
- 9. Memorandum dated May 28, 1997, from John C. Hoyle. Secretary of the )
Commission, to L. Joseph Callan. Executive Director for Operations. NRC. 1
Subject:
Staff Requirements Memorandum- SECY-97-095. "Probabilistic Risk Assessment Implementation Plan Pilot Application for Risk-Informed. Performance-Based Regulation." i l 31
n2
#o UNITED STATES g
8 o NUCLEAR REGULATORY COMMISSION f* E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g g WASHINGTON, D. C. 20556 os, , * ** March 12, 1998 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington. D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
RISK-RANKING APPROACH FOR MOTOR-0PERATED VALVES l During the 449th meeting of the Advisory Committee on Reactor Safeguards (ACRS). March 2-4,1998, we reviewed the efforts of the staff and industry to resolve the issues raised in Generic Letters (GL) 89-10 and 96 05. During this review, we had the benefit of discussions with representatives of the NRC staff and of the documents referenced. This subject has been of continued interest to the ACRS for almost a decade, as evidenced by our report dated May 9,1989, on this subject and by our subsequent periodic reviews of the subject matter. The staff and industry efforts have led to substantial increases in the reliability of motor-operated valves (MOVs) under flow conditions for design-basis accidents. Virtually all licensee responses to GL 89-10 have been closed out and a long-term resolution of this issue will be achieved through GL 96-05. This progress has been made possible, to a large extent, by the cooperation i between the staff and the licensees, as both have recognized the significance of this important safety issue. A Joint Owners Group (J0G). with representation from owners of 93 operating plants, was formed. The J0G proposed a risk-ranking approach for MOVs. The ' frequency of inspection and testing was then related to the risk to plant safety imposed by failure of the individual MOVs. The NRC staff reviewed the proposed l program and issued a Safety Evaluation Report that endorsed the methodology with I some limitations and conditions, which were accepted by the J0G. l l l 33
2 The staff's concern about the emphasis of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code on stroke time and leak testing under normal conditions was resolved by a combination of dynamic testing by the Electric Power Research Institute and instrumented diagnostic tests at nuclear power plants. These results provided an empirical database that is being used to assess the performance of MOVs. Expert panels were used extensively to evaluate the risk ranking of systems and components. The lessons learned from this experience should be helpful to the utilities in the implementation of other programs that rely on expert panels for this purpose. The efforts of the staff and industry have resulted in a program that increases assurance of the proper functioning of MOVs during safety operations. Sincerely, f R. L. Seale Chairman
References:
- 1. NRC Generic Letter 89-10. " Safety-Related Motor-Operated Valve Testing and Surveillance." issued June 28, 1989.
- 2. NRC Generic Letter 96-05. " Periodic Verification of Design-Basis Capability of Safety-Related Motor-0perated Valves." issued September 18.
1996.
- 3. NRC Generic Letter 95-07. " Pressure Locking and Thermal Binding of Safety-Related Power-0perated Gate Valves." issued August 17, 1995.
- 4. V-E-C-1658. Rev.1. " Risk Ranking Approach for Motor-Operated Valves in Response to Generic Letter 96-05." Westinghouse Electric Company. December 1997.
- 5. NRC Safety Evaluation of BWR Owners' Group Topical Report NEDC 32264.
" Application of Probabilistic Safety Assessment to Generic Letter 89-10 Implementation" (Revision 2). February 27, 1996.
- 6. Letter dated October 30,1997. from T. H. Essig. Office of Nuclear Reactor Regulation. NRC. to T. J. Rausch. Commonwealth Edison Company.
Subject:
Safety Evaluation of Joint Owners' Group Program on Periodic Verification of Motor-Operated Valves Described in Topical Report NEDC-32719 (Revision 2). 34
3
- 7. ACRS report dated May 9.1989, fran Forrest J. Remick, Chairman. ACRS.
to NRC Chairman Lando W. Zech, Jr.
Subject:
Generic Letter on Safety-Related Motor-0perated Valve Testing and Surveillance. l 35
/ $a as g'o,, UNITED STATES d c, NUCLEAR REGULATORY COMMISSION M I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS i wAssiworow. p. c. 2osss March 13, 1998 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington. D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
PROPOSED IMPROVEMENTS TO THE SENIOR MANAGEMENT MEETING PROCESS During the 448th and 449th meetings of the Advisory Committee on Reactor Safeguards. February 5-7 and March 2-4. 1998, respectively, we met with representatives of the NRC staff and the Nuclear Energy Institute (NEI) to discuss proposed improvements to the Senior Management Meeting (SMM) process and the efforts of the Integrated Review of Assessment (IRA) Team. Our Subcommittee on Plant Operations discussed these matters during a meeting on February 3,1998. We also had the benefit of the documents referenced. The proposed combined single process is an improvement over the existing three separate but related assessment processes. The new assessment process proposed by the IRA Team will retain many of the positive attributes of the current process. Also, the new process will not preclude taking appropriate regulatory action in a timely manner and will be closely aligned with the NRC's enforcement policy. l Recommendations a We recommend that the documents being developed from the IRA effort not be released for public comment until the staff develops a set of explicit ' program requirements, quantitative if possible, for the plant performance assessment, completes its work on the Assessment Decision Logic Model, and I presents both to the Committee for its review. e The overall objectives stated in Attachment 1 of the draft Commission paper, which was received on February 18. 1998, are not sufficiently specific to allow evaluation of the proposed assessment process. We recommend the development of specific objectives and performance measures 4 that can be applied directly to this process. The Assessment Decision I 37
2 l Logic Model should show how the selected decision options noted in the draft paper will utilize these performance measures. m We recommend that the staff work through at least one example that uses actual inspection reports to demonstrate that the implementation of the Assessment Decision Logic Model is fully understood and workable. This example should include the conversion of the report findings to numerical scores, the processing of these scores through the model, and the decision reached. We would like to review the example before public comments are solicited. m We recommend that the six categories of the proposed template be evaluated to determine that they are at the appropriate level and whether they overlap unnecessarily. This evaluation must be done in the context of the Assessment Decision Logic Model. m We recommend that the staff complete the development and testing of the tools for assessing management and operational effectiveness. The Committee is interested in discussing the results of this efrort with the staff when they have completed their work. m We recommend that economic indicators in their present form not be used in the decisionmaking process at this time and that additional research be performed. m Indicators that measure plant performance at a more global level, such as those discussed by the industry would be more useful. We would like to , see the staff and NEI agree on a set of performance indicators. i a We recommend that the assessment process contain strong provisions to ; ensure that consistent results are obtained among the Regions. l Discussion The Committee has had discussions with the staff and NEI on the status of the NRC Integrated Review of Assessment (IRA) process for operating nuclear power plants. Although the staff has acted upon some previous Committee recommendations. ; additional work remains to be done. As discussed in our September 10. 1997 l report to the Commission, the development of a hierarchical structure of program l requirements and decision logic for the assessment process is important to the l design of the new process. 38
3 In transitioning from a process that had three separate assessments -- systematic assessment of licensee performance (SALP), plant performance review (PPR) and the senior management meeting (SMM) -- to a single assessment process, it is essential to ensure that the requirements of the agency will still be met. These requirements for the single process should be expressed in explicit terms, quantitative if at all possible. A list of these requirements would be useful for evaluating alternate approaches to the assessment process. The staff is assessing the inputs to the Plant Issues Matrix that include most of the licensee performance indicators from the existing assessment process. We believe that these indicators measure performance at such a low level that the nexus between this performance level and overall plant safety is not evident. We believe that the use of indicators that measure performance at a more global level (such as those discussed by the industry) would be more useful. We would like to see the staff and NEI agree to a set of performance indicators. This work could be accomplished during the workshops planned by the NRC staff. At present, the staff has found that economic indicators alone are not useful plant performance indicators. They may have value when used in conjunction with technical plant performance indicators but in their present form are not essential for decisions that have to be made. Because economic pressures arising from deregulation may have a significant effect on long-term safety performance, additional research on economic indicators is needed. The new assessment process moves the evaluation and decisionmaking back to the Regional Offices, where it was before the Senior Management Meeting process began. A key requirement for the new process is that the tools employed, i.e.. the Plant Issues Matrix and Assessment Decision Logic Model. contain provisions to ensure that consistent results are obtained among the Regions. The staff has not completed its work on the Integrated Assessment Process and has not developed an agreed-upon set of requirements for the new process. The process by which the plant performance template leads to the formulation of decisions is not apparent. Development of a hierarchical structure begins with the desired outcome, considers alternate ways to achieve it, and then works down to the most effective means to ensure this outcome. The Committee has yet to see l such a design process applied to this issue. We do not believe the staff will receive useful public comment on the proposed IRA documents as they now exist. We recommend that the documents not be released for public comment until the staff develops a set of requirements for the plant performance assessment i 39
4 program, describes the Assessment Decision Logic Model in sufficient detail, and presents both to the Committee for its review. Sincerely. R. L. Seale Chairman
References:
- 1. Draft Commission paper from L. Joseph Callan, Executive Director for Operations. NRC to the Commissioners,
Subject:
Update on the Status of the Integrated Review of the NRC Assessment Process for Operating Commercial Nuclear Reactors, received February 18. 1998. (Predecisional)
- 2. Draft report (LA-UR-97-4911) dated December 17. 1997, Prepared by Los Alamos National Laboratory for Office of Nuclear Reactor Regulation.
" Integrated Review of the Nuclear Regulatory Commission Assessment Process for Operating Commercial Nuclear Reactors." Working Report 3: Conceptual Design of the Revised Assessment Process. (Predecisional)
- 3. Note dated February 27.1998, from Jack E. Rosenthal, Office for Analysis and Evaluation of Operational Data. NRC. to Michael T. Markley ACRS, transmitting Draft Report AEOD/S98-xx. Prepared by William S. Raughley, AE0D. "Special Study Identifying Financial Indicators." dated February 27, 1998. (Predecisional)
- 4. Memorandum dated January 20,1998. from Richard J. Barrett, AE0D. to John T. Larkins, ACRS, transmitting AEOD draft report. " Interim Report on the Development of the Plant Performance Template." dated January 22, 1998.
(Predecisional)
- 5. Memorandum dated November 6.1997. from C. E. Rossi . AE0D to Addressees.
Subject:
Request for Review of Interim Report - Development and Findings of the Performance Trending Methodology. (Predecisional)
- 6. Memorandum dated February 10, 1998, from John C. Hoyle. Secretary of the Commission, to L. Joseph Callan. Executive Director for Operations NRC.
Subject:
Staff Requirements - Briefing on Operating Reactors and Fuel Facilities. January 21. 1998
- 7. Memorandum dated October 24. 1997 from John C. Hoyle Secretary of the Commission, to L. Joseph Callan, Executive Director for Operations. NRC.
Subject:
Staff Requirements - Briefing on Improvements in Senior Management Assessment Process for Operating Reactors. September 19. 1997.
- 8. Report dated September 10, 1997, from R. L. Seale, Chairman. NRC. to Shirley Ann Jackson, Chairman, NRC.
Subject:
Staff Action Plan to Improve the Senior Management Meeting Process I 40
5
- 9. Memorandum dated September 11. 1997, from John T. Larkins. ACRS. to the Commissioners.
Subject:
ACRS Letter on the Senior Management Meeting Process. September 11, 1997. 1 i I l l 41 i
UNITED STATES l8 NUCLEAR REGULATORY COMMISSION l 'f ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAeNessGTON, D. C.20005
***** March 16,1998 l
Mr. L. Joseph Callan - Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, DC- 20555-0001
Dear Mr. Callan:
SUBJECT:
SECY-98-001. MECHANISM FOR ADDRESSING GENERIC SAFETY ISSUES During the 449th meeting of the Advisory Committee on Reactor Safeguards (ACRS). March 2-4'. 1998, we reviewed the subject SECY paper. During this review, we had l the benefit of discussions with representatives of the NRC staff and of the l referenced documents. The Committee was briefed by the staff on the progress made on Staff Requirements Memorandum (SRM) 951219A concerning mechanisms for addressing generic safety issues (GSIs) Although some progress has been made, the ACRS believes that much work needs to be done to achieve a more efficient prioritization and resolution process. Our principal concerns relate to the following:
- 1. The sporadic issuance of the quarterly reports on the generic information management control system (GIMCS) requires management attention.
GIMCS updates have not been issued in approximately a year, j 2. Several safety-related issues are being tracked in the Office of Nuclear Reactor Regulation (NRR) Director's Status Report but not in GIMCS. If such_ issues in the Director's Status Report are generic in nature, they should be incorporated into the GSI prioritization process and tracked by GIMCS.
- 3. An- adequate- agency: plan or schedule to resolve outstanding GSIs in a timely manner does not exist. Indeed. one HIGH- (GSI-23,1983) and several i MEDIUM-ranked GSIs (GSI B-17.1982: B-55: B-61, 1983) that have been identified for quite some time remain to be resolved. The staff should l establish schedules and resource requirements for resolution of GSIs l according to their prioritization. Issues prioritized as HIGH should not :
remain open for 15 years. j i 43
2 l
- 4. The planning assumptions of the Office of Nuclear Regulatory Research (RES) call for the prioritization of two to three generic issues a year. !
The ACRS has not been briefed on any prioritizations during the past year, i although we have been informed that the staff is currently working on two issues. GSI-169 (Boiling Water Reactor Main Steam Isolation Valve Failure Due to Accumulator Pressure) and GSI-107 (Main Transformer Failures).
- 5. As noted in NUREG/CR-4674. the Office for Analysis and Evaluation of Operational Data reports on operational events indicate that a number of GSIs designated as RESOLVED (e.g., Loss of Offsite Power. GSI-47: Failure of Protective Devices on Essential Equipment. GSI-2: and Containment Emergency Sump Performance. GSI-A43) may not have been adequately resolved.
\
- 6. We also note that agency expectations regarding the resolution of certain GSIs have not always been realized. Certainly, the planned resolution of GSI-172 (Multiple System Responses Program) largely stemmed from the premise that multiple system interaction issues would be addressed in the individual plant examination / individual plant examination of external events (IPE/IPEEE) process. A review of certain IPE/IPEEE submittals, however, has revealed that this premise has not always been correct. The IPEEEs that are based on qualitative analyses such as the Seismic Margins Method and the FIVE (Fire-Induced Vulnerability Evaluation) Method cannot resolve the issues of multiple system interactions.
In view of items 5 and 6. above. the ACRS is concerned about the adequacy ; of the GSI closure process.
- 7. The SRM 951219A encouraged the Office of Nuclear Material Safety and i Safeguards (NMSS) to evaluate and assign priorities to NMSS-related generic issues to ensure consistency of prioritization with reactor-related GSIs administered by RES and prioritized according to risk following the process described in NUREG-0933. In our discussions with 4 the staff, we were informed that the prioritization of nonreactor NMSS-related GSIs was difficult. Even though NMSS has a measure of the consequences associated with a particular event it does not have a method for assessing its frequency. Therefore. NMSS cannot perform a quantitative prioritization process similar to that used for reactor-related GSIs. This circumstance makes the requirement for establishing consistency of prioritization between nonreactor and reactor generic issues a challenge. We encourage NMSS to develop better capability to 44
[ l l 3
- j. apply risk assessment methodology in the prioritization of nonreactor GSIs. Until such capability is developed. the current practice of prioritization on the basis of qualitative consequences should be l continued.
We were informed by the staff that requests are submitted annually to the l regions. NRR, and AE00 to determine if recent operational events warrant reassessment of GSI issues previously classified as LOW in the GSI process. We i recommend that this process be expanded to include GSIs classified as RESOLVED { and that AEOD take the initiative in this regard. The Committee would like to have a briefing from the NRC staff in the near future to discuss plans for resolution of the remaining 15 open GSIs. the process for closure of GSIs and how to handle operational events identified by AEOD that continue to occur after GSIs have been closed. Sincerely. , R. L. Seale Chairman
References:
- 1. SECY-98-001. Memorandum dated January 2.1998. from L. Joseph Callan.
Executive Director for Operations. NRC, for the Commissioners.
Subject:
Staff Requirements Memorandum 951219A - Briefing on Mechanisms for Addressing Generic Safety Issues.
- 2. Memorandum dated January 19. 1996. from John C. Hoyle. Secretary of the Commission. to James M. Taylor. Executive Director for Operations NRC,
Subject:
Staff Requirements - Briefing on Mechanism for Addressing Generic Safety Issues.
- 3. SECY-96-089. Memorandum dated April 30. 1996, from James M. Taylor.
Executive Director for Operations. NRC, for the Comissioners.
Subject:
Comparison of Costs of Generic Requirements Estimated by the NRC with those Estimated by Industry: Staff Effort Expended on Generic Activities.
- 4. SECY-96-107 Memorandum dated May 14, 1996, from James M. Taylor.
Executive Director for Operations. NRC for the Commissioners.
Subject:
Uniform Tracking of Agency Generic Technical Issues. t 45 i
4
- 5. U.S. Nuclear Regulatory Commission. NUREG/CR-4674, Vol. 25. " Precursors to Potential Severe Core Damage Accidents: 1996." A Status Report by ORNL, December 1997.
l
- 6. Memorandum dated February 6,1997, from David L. Morrison, Director. l Office of Nuclear Regulatory Research, NRC, to Addressees,
Subject:
f Periodic Review of Low-Priority Generic Safety Issues. l
- 7. Memorandum dated October 30. 1997. from Carl J. Paperiello, Office of Nuclear Material Safety and Safeguards NRC, to Addressees.
Subject:
NMSS i Policy and Procedures Letter 1-57. Rev.1. "NMSS Generic Issues Program." i i l 2 l l l 46
.[ / %o UNITED STATES NUCLEAR REGULATORY COMMISSION d' ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20556
***** April 9, 1998 Mr. L Joseph Callan Executive Directorfor Opersbons U.S. Nuclear Regulatory Commission Warhington, D.C. 20555-0001
Dear Mr. Callan:
SUBJECT:
THE SAFETY ASPECTS OF THE WESTINGHOUSE ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE AP600 PLANT DESIGN - INTERIM LETTER 2 During the 451st meeting of the Advisory Committee on Reactor Safeguards, April 2-4,1998, we reviewed various chapters of the AP600 Standard Safety Analysis Report (SSAR) and associated chapters of the draft Final Safety Evaluation Repcrt (FSER). Our Subcommittee on Advanced Reactor Designs reviewed these chapters on March 31-April 1,1998. During these reviews, we had the benefit of discussions with representatives of the NRC staff and the Westinghouse Electric Company and of the documents referenced. We reviewed the following SSAR and associated draft FSER chapters:
. Chapter 2 - Site Charactensbcs ; - Chapter 9 - Auxiliary Systems, including Appendix 9A- Fire Protechon Analysis ) . Chapter 10 - Steam and Power Conversion System . Chapter 12 - Radiabon Protechon . Chapter 13 - Sechon 13.6, industnal Security i Chapter 15 - Accident Analyses Based on our review of the above SSAR and as:ociated draft FSER chapters, we offer the following comments and recommendsbons.
Chapter 2 - Site Charactensbes in order to meet the 10 CFR 50.34, " Contents of applications; technical information," siting dose criteria, site-specific short-term and long-term atmospheric dispersion factors X/ Q should meet the AP600 design acceptance values. These acceptance values place limits on the outcomes of site-specific atmospheric dispersion calculations and define bounds on the meteorological characteristics of acceptable sites. The NRC staff should ensure that the calculatione.1 methodologies used by the Combined License applicant to derive x/Q not mask the effects of any unique site meteorological 47
l characteristics related to topology, geographical location, directed wind flows during specific ! times of the day, or any peculiar atmospheric inversion characteristics. l I Chanter 9 - Auviliary Systems. Includina Anoendir 9A - Fire Pie 6ctier, Analvais I L We have not completed our review of the fire protection system and the fire protection analysis. We plan to provide our views at a later date. Chapter 13 - Section 13.6. Industrial Security 1 The design simplicity of the AP600 permits a security design that eliminates the need for some of the features described in 10 CFR 73.55, " Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage." The number of access portals to the vital areas is reduced to a minimum to accommodate personnel and material flow. This approach enhances the security design and requires less manpower for responding to attempted intrusions. Although the proposed security design meets the applicable regulations, it appears to make routine operations and maintenance activities more difficult. Restrichons such as personnel searches will cause delay for personnel when entering areas of the plant containing equipment, which must be surveilled, operated, and maintained. These restrictions could have safety implications, especially during plant upset conditions, emergencies such as fires, and implementation of the station Emergency Plan. We recommend that the staff evaluate the impact of the security design on the safety aspects of the plant. The additional time required to perform personnel searches might well increase the expected response times of the fire brigade and operators attempting to mitigate safety significant events. We expect to review the results of this evaluation. Chaoter 15 - Accident Analyses in assessing the acceptability of the AP600 design in meeting the regulatory dose criteria associated with design-basis accidents (DBAs), Westinghouse has propcsed taking credit for natural aerosol removal processes that occur in the containment. These processes include agglomeration, gravitational sedimentation, diffusional plateout, diffusiophoresis, and thermophoresis. In the past, the design-basis source term was considered to account implicitly for these and other natural aerosol removal processes in both the reactor coolant system and in , the containment. The new source term released to the containment implicitly accounts for only ! the effects of aerosol removal processes in the reactor coolant system. Radionuclide l deposition in the containment must now be calculated exphcstly. The processes of agglomeration, sedimentation, and diffusion are present at all times for all accident sequences and containment designs (and are not significantly dependent on thermal-hydraulic conditions). Thus, these specific processes can be accounted for explicitly without specifying the thermal-hydraulic conditions. The processes of diffusiophoresis and thermophoresis, on the other hand, depend on the synchronization of the thermal-hydraulic 48
processes (steam condensation rates and development of thermal gradients across boundary layers) with the associated source-term concentrations in the containment. For current operating plants, the relative timing of source term release and thermal-hydraulic phenomena have been shown to be dependent on both plant design and event sequence. It is not clear what thermal-hydraulic conditions should be associated with the DBA source term, which is an cmalgamation of source terms associated with a range of severe acculent sequences. Westinghouse has chosen the thermal-hydraulic conditions of a specific sequence (i.e., a direct vessel injection line break) for use with the DBA source term to take credit for diffusicit.ciesis End thermophoresis. The DBA concept is intended to ensure that the containment design results in an acceptable risk for all accident sequences. It is not clear that the thermal-
- hydraulic conditions of the selected sequence is consistent with the desired generality of the j source term. i l l Specifying the thermal-hydraulic conditions associated with a specific sequence for use with the I new source term appears to constitute an unprecedented interpretation of the design-basis )
concept. We recommend that the justification for granting credit for aerosol removal due to diffusiophoresis and thermophoresis be documented. Furthermore, the staff should make it clear that such credit is not intended to be generic for other plant designs. Dr. Dana A. Powers did not participate in the Committee's deliberations regarding the DBA source term. l Sincerely, l A
/f. . sv-lA R. L Seale Chairman References-
- 1. Letter dated February 19,1998, from R. L Seale, Chairman, ACRS, to L Joseph l Callan, Executive Director for Operations, NRC,
Subject:
Interim Letter on the Safety l Aspects of the Westinghouse Electnc Company Application for Certification of the l AP600 Plant Design. I
- 2. U.S. Department of Energy report prepared by Westinghouse Electnc Corporation, ,
! "AP600 Standard Safety Analysis Report," updated through Revision 20 dated February 6,1998.
- 3. Memorandum dated March 9,1998, from Jack W. Roe, Office of Nuclear Reactor Regulation, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
Transmittal of Advanced Copy of Chapter 12 of the AP600 Final Safety Evaluation Report (FSER) (Predecisional Draft). 49
- 4. Memorandum dated March 17,1998, from Jack W. Roe, Office of Nuclear Reactor Regulation, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
Transmittal of Advanced Copy of Chapter 2 of the AP600 Final Safety Evaluation Report (FSER) (Predecisional Draft).
- 5. Memorandum dated March 17,1998, from Jack W. Roe, Office of Nuclear Reactor i Regulation, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
Transmittal of Advanced Copy of Chapter 9 of the AP600 Final Safety Evaluation Report (FSER) (Predecisional Draft). i
- 6. Memorandum dated March 17,1998, from Jack W. Roe, Office of Nuclear Reactor l Regulation, NRC, to John T. Larkins. Executive Director, ACRS,
Subject:
Transmittal of ! Advanced Copy of Chapter 10 of the AP600 Final Safety Evaluation Report (FSER) (Predecisional Draft).
- 7. Memm xium dated March 17,1998, from Jack W. Roe, Office of Nuclear Reactor i
RegFc* 1, NRC, to John T. Larkins, Executive Director, ACRS, Subject Transmittal of Advs. . ,d Copy of Chapter 15 of the AP600 Final Safety Evaluation Report (FSER) j (Predecisional Draft).
- 8. Memorandum dated March 24,1998, from Jack W. Roe, Office of Nuclear Reactor Regulation, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
Transmittal of Advanced Copy of Section 9.5.1 of the AP600 Final Safety Evaluation Report (FSER) I (Predecisional Draft).
- 9. Memorandum dated March 23,1998, from Jack W. Roe, Office of Nuclear Reactor Regulation, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
Transmittal of Advanced Copy of Section 13.6 of the AP600 Final Safety Evaluation Report (FSER) (Predecisional Draft).
- 10. Westinghouse AP600 Security Design Vulnerability Analysis Report, GS-ASR-002, ,
Revision 2, dated February 3,1998 (Safeguards information). '
- 11. Westinghouse AP600 Security Design Report, GS-ASR-001, Revision 5, dated February ,
3,1998 (Safeguards information). t 50
maag% UNITED STATES
/ o NUCLEAR REGULATORY COMMISSION
$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ej, WASHINGTON, D. C. 20086
..... April 9, 1998 Mr. L Joseph Callan Executive Drector for Operabons U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Mr. Callan:
l
SUBJECT:
Pl.ANS TO INCREASE PERFORMANCE-BASED APPROACHES IN REGULATORY ACTIVITIES During the 451st meebng of the Advisory Committee on Reactor Safeguards, April 2-4,1998, we
- met with representatives of the NRC staff to discuss the proposed Commission paper entitled
' Plans to increase Performance-Based Approaches in Regulatory Activities." Our Subcommittee i
on Reliability and Probabilistic Risk Assessment met with the staff on November 13,1997, and l February 20,1998, to discuss issues related to performance-based regulation. We also had the benefit of the documents referenced. The proposed Commission paper provides early input to the dQTei.i of Strategy 5 of the NRC Excellence Plan. Strategy 5 is intended to develop a process and identify candidate issues for improvmg the effeebveness and effiaency of rules, regulatory guidance and their application. We i believe that the staff should identify and define the issues likely to arise in developng performance-I based regulations in the absence of quantified risk information. Examples include: how and by whom performance parameters are to be determined and deemed acceptable, and how to place the acceptance limits on them. A clearer description of the issues at this time would help structure the proposed solicitation of input from the public and interested stakeholders. We recommend that the staff provide the above information to us prior to proceeding with the solicitation of input. We look forward to working with the staff as it develops the strategy. Sincerely, I . sr
/J. -
R. L Seale Chairman References-
- 1. Draft Commission paper for the Commissioners from L Joseph Callan, Executive Director i for Operations, NRC, Subject Plans to increase Performance-Based Approaches in l Regulatory Activities, received March 25,1998 (Predecisional).
- 2. Draft Commission paper for the Commissioners from L Joseph Callan, Executive Director for Operations, NRC,
Subject:
FY 1998 NRC Excellence Plan, received March 13,1998 (Predecisional). 51
#g UNITED STATES / ,,
NUCLEAR REGULATORY COMMISSION y, I ADVISORY COMMITTEE ON REACTOR SAFESUARDS g wAsumarow, p. c.zoses April 9, 1998 MEMORANDUM TO: L. Joseph Callan Executrve Director f+=.1- - FROM:
. [
John T. Larkins, ecutive Director Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED RULEMAKING: CHANGES TO PARAGRAPH (h) OF 10 CFR 50.55a, " CODES AND STANDARDS" During the 451st meeting of the Advisory Committee on Reactor Safeguards, April 2-4, 1998, the Committee decxled to consider the proposed final changes to paragraph (h) of 10 CFR 50.55a after reconciliation of public comments Reference-Memorandum dated March 13,1998, from Samuel J. Collins, NRR, to John T. Larkins, Executive Director, ACRS, and Thomas T. Martin, Chair, CRGR,
Subject:
Proposed Rulemaking: Changes to Paragraph (h) of 10 CFR 50.55a, " Codes and Standards." cc: J. Hnyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR T. Martin, AEOD M. Knapp, RES f 53
** aso ug'o, UNITED STATES
/ o NUCLEAR REGULATORY COMMISSION $ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ' R WASHINGTON, D. C. 20655 g, o.. . . */- April 9, 1998 MEMORANDUM TO: L Joseph Callan Executive Direp9r%. FROM: John T. Larkins, ec e irector Advisory Committee on Reactor Safeguards
SUBJECT:
DRAFT COMMISSION PAPER, " STATUS OF THE ENVIRONMENTAL QUALIFICATION TASK ACTION PLAN" During the 451st meeting of the Advisory Committee on Reactor Safeguards, April 2-4, 1998, the Committee decided not to review the subject draft Commission paper. The Committee, however, plans to review the proposed resolution of Generic Safety issue 168,
" Environmental Qualification of Electrical Equipment," when it is available.
References
- 1. Draft memorandum received February 26,1998, from L. Joseph Callan, Executive Director for Operations, NRC, to NRC Chairman and Commissioners,
Subject:
Status of the Environmental Qualification Task Action Plan (Predecisional).
- 2. Office of Nuclear Reactor Regulation, " Environmental Qualification Task Action Plan,"
dated September 2,1997, regarding GSI 168, " Environmental Qualification of Electrical Equipment," Revision 1 dated June 30,1995 cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR G. Holahan, NRR G. Hubbard, NRR T. Martin, AEOD M. Knapp, RES 55
E
/ 's,
'^r, UNITED STATES l 8 NUCLEAR REGULATORY COMMISSION f
I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsHWGTON. D. C. 20555
***** April 9,1998 1
MEMORANDUM TO: L. Joseph Callan Executive Di '- -J.=2-r FROM: John T. Larkins, Di or Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED FINAL GENERIC LETTER, " YEAR 2000 READINESS OF COMPUTER SYSTEMS AT NUCLEAR POWER PLANTS" During the 451st meeting of the Advisory Committee on Reactor Safeguards, April 2-4, 1998, the Committee decided not to review the subject Generic Letter. Reference-Memorandum from Frank J. Miraglia, Jr., Office of Nuclear Reactor Regulation, NRC, to i Thomas T. Martin, Chairman, Committee to Review Generic Requirements, NRC,
Subject:
Request for Review and Endorsement of the Proposed Generic Letter Entitled " Year 2000 R:adiness of Computer Systems at Nuclear Power Plants," received April 2,1998 (Predecisional Draft). j cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR B. Boger, NRR T. Martin, AEOD M. Knapp, RES A. Galante, CIO l 57 l
#g UNITED STATES
/ NUCLEAR REGULATORY COMMISSION y ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHNGTON, D. C. 20555 April 9,1998 MEMORANDUM TO: L Joseph Callan Executive Director for FROM: John T. Larkins, Ek' ACRS/ACNW
SUBJECT:
DRAFT REGULATORY GUIDE DG-1078, " STANDARD FORMAT AND
- CONTENT OF LICENSE TERMINATION PLANS FOR NUCLEAR POWER REACTORS" ,
During the 451st meeting of the Advisory Committee on Reactor Safeguards, April 2 4, 1998, the Committee considered the proposed final subject Regulatory Guide and transferred review of this guide to the Advisory Committee on Nuclear Waste. The ACNW will consider the proposed final Regulatory Guide after reconciliation of public comments and wishes to receive a general bnefing on the regulatory process for power reactor decommissioning and an update on current reactor decommissioning and decontamination plans. Reference-Draft Regulatory Guide DG-1078, " Standard Format and Content of License Termination Plans for Nuclear Power Reactors," March 1998. l oc- J. Hoyle, SECY J. Blaha, OEDO . J. Mitchell, OEDO C. Pittiglio;NMSS J. Hickey, NMSS 59
/ g UNITED STATES
/ n NUCLEAR REGULATORY COMMISSION $ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20556 May 8,1998 Mr. L. Joseph Callan Executive Drector for Operatens U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Mr. Callan:
SUBJECT:
PROPOSED FINAL AMENDMENT TO 10 CFR PART 55, "lNITIAL LICENSED OPERATOR EXAMINATION REQUIREMENTS" During the 452nd meeting of the Advisory Committee on Reactor Safeguards, April 30 - May 2, 1998, we met with representatrves of the NRC staff and the Nuclear Energy institute to discuss the proposed final amendment to 10 CFR Part 55 regarding initial licensed operator examinsbon requirements We also had the benefit of the document referenced. RECOMMENDATIONS:
- 1. We recommend that the amendment to 10 CFR Part 55, " Initial Licensed Operator Examinsbon Requirements," be issued for use by the industry.
- 2. We recommend that the sta# analyze results of the 68 pilot examinations to ensure that the quality and level of difficulty of the examinations are consistent across the regions.
DISCUSSION Transfer of the responsibility for preparing and administering initial licensed operator Examinations to licensees may reduce confidence that operators will place safe plant cperations above other imperatives imposed by plant management. There is a concem that there could be potential shifts in the obligations felt by those taking the examination. To guard against this, performance measures on the effectiveness of the examination process should be developed and assessed periodmally. We believe that the benefits of the new process far outweigh the potential drawbacks. This ; process will provide more assurance that plant-specific aspects of the examinations are j accurate and up to date. Indeed, the NRC staff will retain its authority to perform appropriate checks on operator qualifications and will continue to directly observe and evaluate the performance of every license applicant on the operating tests (simulator and job performance measures). The NRC staff obligations to ensure adequacy and accuracy of examinations prepared by licensees will remain the same. Results of the NRC staffs review of the pilot program showed that the quality of the cxaminations varied widely and that many examinations needed significant revision. Therefore, 61
2 \ ! the NRC staff concluded that the quality and level of difficulty of the examinations require continued attention. Our understanding is that this review was not done across the regions to ensure consistency in the quality and level of difficulty. l l The new examination process makes the overall operator licensing program more consistent l with the Agency's other oversight programs. It holds the licensee accountable for the quality of the examinations. The NRC examiners may now focus more on the cognitive level of the questions and the plausibility of the distractors (wrong-answer choices). We believe that the results achieved and the experience gained from the pilot program have shown that the preposed amendment to 10 CFR Part 55 can result in an effective examination ) process. I Sincerely, R. L. Seale L Chairman l l Reference l Memorandum dated April 20,1998, from Jack W. Roe, Office of Nuclear Reactor Regulation, to John T. Larkins, Advisory Committee on Reactor Safeguards,
Subject:
Transmittal of Advanced Copy of the Final Rule on Requirements for Initial Operator Licensing Examinations 4 (10 CFR Part 55)(Predecisional Draft) 1 62
# 'o,, UNITED STATES
/ o NUCLEAR REGULATORY COMMISSION $ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Q WASHINGTON, D. C. 20555 May 8,1998 MEMORANDUM TO: L. Joseph Callan Executrve Di - FROM: John T. Larkins, Exe [Sn m_ x e e Director Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED GENERIC LETTER, " MODIFICATION OF THE NRC STAFF'S RECOMMENDATIONS FOR THE POST-ACCIDENT SAMPLING SYSTEM" During the 452nd meeting of the Advisory Committee on Reactor Safeguards, April 30-May 2,1998, the Committee decided not to review the subject generic letter.
Reference:
Proposed NRC Generic Letter 98-YY, " Modification of the NRC Staffs Recommendations for the Post-Accident Sampling System," received April 14,1998 (Predecisional Draft). cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR G. Lainas, NRR T. Martin, AEOD M. Knapp, RES I i l 63
i sen , UNITED STATES 8 o,% NUCLEAR REGULATORY COMMISSION $ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASMNGTON, D. C. 20555 May 11,1998 The Honorable Shirley Ann Jackson Chairman U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
ELEVATION OF CDF TO A FUNDAMENTAL SAFETY GOAL AND POSSIBLE l REVISION OF THE COMMISSION'S SAFETY GOAL POLICY STATEMENT During the 449th,451st and 452nd meetings of the Advisory Committee on Reactor Safeguards, March 2-4, April 2-4, and April 30-May 2,1998, respectively, we met with representatives of the NRC staff and the industry to discuss the elevation of core damage frequency (CDF) to a fundamental safety goal and the need for revising the Commission's Safety Goal Policy Statement. In our August 15,1996 report to the Commission, we recommended elevating CDF to a fundamentalsafety goal. Our Subcommittee on Reliability and Probabiliste Risk Assessment met on February 20 and April 16,1998, to discuss these matters. Former NRC Commissioner Forrest Remick and former ACRS Member David Okrent participated in these discussions. We also had the benefit of the documents referenced. The Quantitative Health Objectives (QHOs) regarding individual risk and societal risk promulgated in the Safety Goal Policy Statement are fundamental goals. Due to the large uncertainties in evaluatingindividualand societal risk, attemative objectives are often used. The most frequently used attematives are limits on CDF and on large, early release frequency (LERF). These are generally referred to as " subsidiary" or " surrogate" goals to indicate that they are intended to be consistentwith the fundamental safety goals (to the extent that the cunrent computationalcapability allows the determinationof consistency'). If one of these numerical goals were to be significantly more conservative than the OHOs, then it would no longer be subsidiary, but could itself be a new, , de facto fundamental safety goal. in its report dated May 13,1987, the ACRS discussed a hierarchical structure to facilitate the implementation of the Safety Goal Policy Statement, and expressed concem over a de facto iln this regard, the Policy Statement advises that "It is the Commission's intent that the risks from all the various initiating mechanisms be taken into account to the best of the capability of current evaluation techniques." The check for consistency to which we are referring will also have to be done using the best current evaluation techniques. 65
i 2 change in policy: "Each subordinate level of the hierarchy should be consistent with the level above, but should be a more practical surrogate, representing a simplification or quantification of the previous level. Each surrogate should not be so conservative that it creates a de facto new policy." The staff has stated numerical objectives for CDF and LERF in Regulatory Guide 1.174 and has employed these values in other regulatory decisionmaking. The question is whether these numericalvalues are consistentwith the QHOs. There are plants that could have CDF values even 3 greater than 104 per reactor-year and still be consistent with the QHOs because of their containment performance. Thus, we make the following observations: . l Observahon.1. Existing Level-3 PRA results can be used to estimate specific values for LERF that we d be consistentwith the QHOs. Such estimates show that the current LERF obective l of 10 4per reactor-year is consistent with the QHO on early fatalities and, I therefore, could be considered a surrogate goal. Observation 2 Results of analysesindicatethat a CDF objective of 104 per reactor-year, if applied to all plants with their current level of containment performance, in many cases would be more conservative than the QHOs. This would, therefore, be a new de facto fundamental safety goal. Although we agree with the criteria on CDF and LERF included in Regulatory Guide 1.174, we believe that such de facto changes in policy are not desirable. The issue of elevating CDF to a fundamental safety goal should be thoroughly scrutinized. irrespective of Observation 2, we agree with the staff position that a revision of the Policy . Statement is needed to address the use of the goals on a plant-specific basis, to expand the discussion on uncertainties, and to remove the general plant performance guideline. These and otherissues regarding the current Policy Statement should be addressed if a revision is considered. Even though the Policy Statement refers to societal risk, the application guidance and practice resultin essentially individualrisk goals. Furthermore, environmental contamination and the total number of fatalities have been mentioned often as being appropriate societal measures of the consequences of accidents at nuclear power plants. Environmental contamination receives close attention within the nuclear regulatory framework of some other countries. The importance of environmental contamination resulting from accidents is recognized by the NRC staff in regulatory analyses and in environmental impact statements. The question is whether this and the total number of fatalities should be part of the Safety Goal Policy. 1 Thus, we make the following recommendation: l Recommendation 1* There is a need to revise the Safety Goal Policy Statement. The revision should include: (a) a statement regarding the plant-specificuse of the safety goals; (b) an expanded treatment of the role of uncertainties; (c) the removal of the general plant performanceguideline; (d) a reconsiderationof the set of fundamentalgoals and subsidiary objectives to ensure that they are consistent; and (e) a reconsideration of measures of societal risk such as environmental contamination and the total number of fatalities. l 66
3 During our discussions with the staff, a question was raised regarding the level of detail that the Policy Statement should contain. Should such a document only express the overall safety policy and regulatory approach of the Agency, or should it also speedy numerical values for the metrics? The current Statement does specify numerical QHOs. If it is decided to elevate CDF to the same level, should it simply state that the prevention of core damage is a fundamental safety goal and leave the specification of actual numerical guidelines in application guidance? In either case, attention will be drawn to the significance that the Commission places on core damage accidents end would be a clear statement of defense-in-depth in terms of PRA. We understand the attractiveness of providing clear, understandable criteria, yet the inclusion of too many quantiative objectives might be overly constraining and not sufficiently flexible to adapt to changes in technology. An additionalimportant conceptual issue is whether the objectives should be stated in terms of a single goal or a goal and an upperlimit. The current Policy Statement specifies only a single goal for each objective, e.g., if the calculated risk of prompt fatality to an individual in the vicinity of a nuclear power plant is less than 0.1 percent of the sum of prompt fatality risks from other accidents, then that plant meets this objective, otherwise it does not. An upper limit and a goal define three regions. For risk levels above the upper limit, immediate tction should be taken. For risk levels between the upper limit and the goal, the possibility of reducing the estimated metric should be investigated, taking into account costs and benefits. For risk levels below the goal, no action would be required. This approach would be consistent with the " risk-informed" philosophy, which recognizes that risk metrics are only part of the decisionmaking process, but if the value of a risk metric were found to be very large, this would i lead to immediate action. The use of two values for making decisions involving risk metrics has been adopted by the nuclear regulatory agencies of The Netherlands and the United Kingdom (Versteeg,1992; Ballard,1993). Similarly,in a report dated October 31,1980, the ACRS recommended decision rules that would cmploy a " goal level" and an " upper limit" on the various frequencies of the risk metrics. Even though the Commission did not adopt this earfier ACRS recommendation as part of its Safety Goal Policy Statement, it appears that both the staff and the industry act as if it were in effect. They s respond immediately when a contributor to core damage is identified that increases the CDF to cbout 104 per reactor-year or greater. Some examples illustrating this behavior are the discoveries, by Individual Plant Examirations (IPEs) and Individual Plant Examination of Extemal Events (IPEEEs), of such potential contributorsinitiated by intemal flooding at the Surry plant and by fire at the Quad Cities plant. The recently published report on the IPE program (NUREG-1560) states that "the CDFs for all boiling water reactors (BWRs) and most pressurized water reactors (PWRs) fall below 1E-4/ry; however, nine licensees representing 15 PWR units reported CDFs above 1E-4/ry" (page 7-3). Given the limited scope of the IPE studies, it is reasonable to expect that the number of units (both PWR and BWR) with CDFs greater than 10d per reactor-year is higher than the IPE findings. Thus, if the CDF value of 104 per reactor-year were to become a fundamental safety goal, the two-intervalapproach might lead to the perception by members of the public that the units having 67 , 1
4 CDFs greater than the goal are " unsafe." In a letter dated July 23,1997, to NRC Chairman Jackson, Mr. J. F. Colvin of the Nuclear Energy Institute articulated this concern: " using core damage frequency as a fundamental safety goal now would send a message to the public that plants that exceed the core damage frequency objective are unsafe, even though they may be well below the safety goal quantitative health objectives." The three-region formulation helps to alleviate this problem. Therefore, we offer the following recommendation: Recommendation 2: If revised, the Policy Statement should be written in terms of high-level principles and expectations and should include numerical guidelines on fundamental goals. We continue to believe that CDF should be elevated to a fundamental safety goal. Using three regions for some of the objectives should be evaluated, as opposed to the two that the current Policy Statement identifies. We believe that a revision of the Policy Statement as discussed above would be a major undertaking. The staff stated that revising the Safety Goal Policy Statement would necessitate the reallocation of limited staff resources and would have an adverse impact on risk-informed regulatory activities. We view the completion and implementation of the Standard Review Plan and Regulatory Guides associated with risk-informed regulation as having great and immediate importance. We are, thus, led to the following recommendation: Recommendation 3. The staff's request to defer modifying the Policy Statement for one year to permit evaluation of related issues and impacts should be approved. We plan to continue our discussions with the staff regarding these matters. 4 Sincerely,
. . A R. L. Seale Chairman References
- 1. Memorandum dated October 16,1997, from John C. Hoyle, Secretary, to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
Staff Requirements - SECY-97-208,
- Elevation of the Core Damage Frequency Objective to a FundamentalCommission Safety Goal."
- 2. Memorandum from L. Joseph Callan, Executive Director for Operations, NRC, for the Commissioners,
Subject:
" Modifications to the Safety Goal Policy Statement." received March 26,1998 (Predecisional Draft).
- 3. Report dated May 13,1987, from William Kerr, Chairman, ACRS, to Lando W. Zech, Jr.,
Chairman, NRC.
Subject:
"ACRS Comments on an implementation Plan for the Safety Goal Policy."
68
5
- 4. Report dated August 15,1996, from T. S. Kress, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
" Risk-Informed, Performance-Based Regulation and Related Matters."
- 5. SECY-98-015, Memorandum dated January 30,1998, from L. Joseph Callan, Executive Director for Operations, NRC, for the Commissioners,
Subject:
" Final General Regu'atory Guide and Standard Review Plan for Risk-informed Regulation of Power Reactors."
(Predecisional Draft).
- 6. M. F. Versteeg, " Showing Compliance with Probabilistic Safety Criteria and Objectives,"
Reliability Engineering and System Safety, 35 (1992) 39-48.
- 7. G. Ballard,"Guesteditorial: SocietalRisk-Progresssince Farmer,"ReliabilityEngineering and System Safety, 39 (1993) 123-127.
- 8. Report dated October 31,1980, from Milton S. Plesset, Chairman, ACRS, to John F.
Aheame, Chairman, NRC,
Subject:
"An Approach to Quantitative Safety Goals for Nuclear Power Plants."
- 9. U. S. Nuclear Regulatory Commission, NUREG-1560, " Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance," Vols.1-3, December 1997,
- 10. Letter dated July 23,1997, from J. F. Colvin, President and CEO, Nuclear Energy Institute, to Shirley Ann Jackson, Chairman, NRC, regarding elevation of the core damage frequency subsidiary objective to a fundamental safety goal.
- 11. Memorandum dated July 2,1997, from Shirley Ann Jackson, Chairman, NRC, to L. Joseph Callan, Executive Directorfor Operations, NRC,
Subject:
"The Statement of Core Damage Frequency of 1E-4 as a Fundamental Commission Goal."
- 12. Memorandum dated October 8,1997, from NRC Commissioner Diaz, to John T. Larkins, ACRS,
Subject:
" Safety Goal."
- 13. U. S. NuclearRegulatoryCommission, Policy Statement,"SafetyGoalsforthe Operations of Nuclear Power Plants," 10 CFR Part 50, August 21,1986. i 1
69
o uru UNITED STATES 8 % NUCLEAR REGULATORY COMMISSION $ E AovisORY COMMITTEE ON REACTOR SAFEGUARDS $ WASHWGTON, D. C. 20555 May 12,1998 MEMORANDUM TO: L Joseph Callan Executive Directorfor Operatums FROM: John T. La Executiv rect Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED HUMAN PERFORMANCE PLAN AND PROPOSED FINAL STANDARD REVIEW PLAN SECTION AND REGULATORY GUIDE FOR RISK-INFORMED INSERVICE INSPECTION OF PIPING During the 452nd meeting of the Advisory Committee on Reactor Safeguards, April 30 - May 2,1998, the Committee held discussions with the NRC staff conceming the subject documents. Since the Committee did not receive these documents in time for sufficent review, its comments will be available after the June 3-5,1998 ACRS meeting. References *
- 1. Memorandum dated April 23,1998, from T. L. King, Office of Nuclear Regulatory Research, to J. T. Larkins, ACRS,
Subject:
NRC's Human Performance Plan
- 2. Memorandum dated April 28,1998, from M. W. Hodges, Office of Nuclear Regulatory Research; G. M. Holahan, Office of Nuclear Reactor Regulation; G. C. Lainas, Office of Nuclear Reactor Regulation, to J. T. Larkins, ACRS,
Subject:
Regulatory Guidance Document [RG 1.178 - (Formerly DG-1063)] and Standard Review Plan (Section 3.9.8) for Risk-informed Inservice inspection of Piping cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR G. Holahan, NRR G. Lainas, NRR M. Knapp, RES T. King, RES M. W. Hodges, RES 71
UNITED STATES 8 NUCt. EAR REGULATORY COMMISSION $ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS e WASHINGTON, D. C. 20505 June 9,1998 Mr. L Joseph Callan Exec %3 Director for Operations U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Mr. Callan:
SUBJECT:
NRC PARTICIPATION IN THE CABRI REACTOR FUELS RESEARCH PROGRAM During the 453rd meeting of the Advisory Committee on Reactor Safeguards, June 3 5,1998, we discussed with representatives of the NRC staff the proposed plans for the NRC reactor fuels research program. Our Subcommittee on Reactor Fuels also discussed this matter during a meeting on April 23-24,1998. We also had the benefit of the document referenced. The regulatory requirements for design-basis safety analyses require consideration of the ability of reactor fuel to sustain energy inputs caused by reactivity insertion events. Current regulatory guidance indicates fuel can tolerate energy inputs of 180 to 280 cal /g regardless of the level of fuel bumup. Results of the experiments performed over the last few years have revealed that the cbility of fuel to sustain sudden energy inputs decreases substantially with increasing fuel bumup. Among the test facilities used to perform experiments that have provided these findings is the French CABRl reactor, which now has a sodium experimental loop. Revised limits on tolerable energy inputs to reactor fuel by reactivity insertion are needed for high bumup. Establishing revised limits based on empirical data is difficult because the available database is small. Furthermore, some data obtained for high bumup fuels have been questioned because of nonprototypic features of t% tests. Clearty, a more extensive, better database would be useful. The Office of Nuclear Regulatory Research (RES) proposes to participate in upgrading the CABRI reactor to include a water loop for experiments. The upgraded reactor could then be used to obtain additional data on fuel response to reactivity insertion. RES has asked for our cdvice on the proposed participation in the CABRI upgrade. Participation in the CABRI research program is a leveraged mechanism for obtaining integral test data on the response of high bumup fuels with modem cladding to reactivity insertion events. The CABRI reactor, when equipped with a water loop, will be the worid's best test facility for obtaining these data, integral data of this type are needed to maintain an adequate technical foundation for NRC regulatory guidance on reactor fuel performance. We concur with the staffs i recommendation to participate in this collaborative research program. J 73 i
2-We recommend that the staff use quantitative methods to establish its positions on test matrices for the CABRI program. Quantitative experiment design methods should be used to determine the need for replicate tests and to determine the accuracy and precision of test data required to make substantive improvements to the existing base of knowledge. In October 1998, we would like to discuss the status of the NRC participation in the CABRI program. Dr. William J. Shack did not participate in the Committee's deliberation regarding this matter. Sincerely,
. 4 , -y -- 4
, Y R. L. Scale Chairman
Reference:
Memorandum dated May 22,1998, from Thomas L King, RES, to John T. Larkins, Executive Director, ACRS,
Subject:
Transmittal of Advance Copy of Agency Program Plan for High-Bumup Fuel (Predecisional) 74
Itte UNITED STATES 8 i ' NGCLEAR REGULATORY COMMISSION. d ADVISORY C0hMITTEE ON REACTCR SAFE 2UARDS o , . - . WASHINGTON, D. C. 30806
%, o %.gwl sr** '
ose* June 9,1998 MEMORANDUM To: L JosephCallan Executive or s __ ,v FROM: John T. Larkins Director Advisory Committee on Reactor Safeguards
SUBJECT:
SAFETY EVALUATION REPORT ON ELECTRIC POWER RESEARCH INSTITUTE TOPICAL REPORT TR-107330, FINAL REPORT," GENERIC REQUIREMENTS SPECIFICATION FOR QUALIFYlNG A COMMERCIALLY AVAILABLE PLC FOR SAFETY-RELATED APPLICATIONS IN NUCLEAR POWER PLANTS " During the 453rd meeting of the Advisory Committee on Reactor Safeguards, June 3-5,1998, the Committee decided that it has no objection to issuing the subject report. Reference-Memorandum dated May 22,1998, from Frank J. Miraglia, Office of Nuclear Reactor Regulation,
' to Thomas T. Martin, CRGR, NRC,
Subject:
Request for Endorsement of the Safety Evaluation Report on Electric Power Research Institute Topical Report, TR-107330, Final Report, " Generic Requirements Specification for Qualifying A Commercially Available PLC for Safety-Related Applications in Nuclear Power Plants." cc: J. Hoyle SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR L. Spessard, NRR R. Gallo, NRR T. Martin, AEOD M. Knapp, RES - A. Galante, CIO 75
asc oq'o,
, UNITED STATES ! n NUCLEAR REGULATORY COMMISSION U I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS waswincrow.o. c.zo!uis June 9, 1998 MEMORANDUM T0: L. Joseph Callan Executive Director ons FROM: John T. Larkins. xe 1 e Director Advisory Comittee on Reactor Safeguards
SUBJECT:
DRAFT ADVANCE NOTICE OF PROPOSED RULEMAKING REGARDING AMENDING 10 CFR 50.72. "IMMEDIATE NOTIFICATION REQUIREMENTS FOR OPERATING NUCLEAR POWER REACTORS." AND 10 CFR 50.73. LICENSEE EVENT REPORT SYSTEM
- During the 453rd meeting of the Advisory Comittee on Reactor Safeguards. June 3-5. 1998. the Comittee decided not to review the subject document. and has no objection to issuing it for public coment. The Comittee, however, would like the opportunity to review the proposed rule when it becomes available.
Reference:
U. S. Nuclear Regulatory Comission. Draft Advance Notice of Proposed Rulemaking.10 CFR Part 50 " Event Reporting Requirements for Nuclear Power Reactors." received June 2. 1998. cc: J. Hoyle. SECY J. Blaha. OEDO J. Mitchell. OEDO T. Martin. AE00 P. Baranowsky. AE00 l 0. Allison. AE00 77 l
#pm asog'o,, UNITED STATES 8
o NUCLEAR REGULATORY COMMISSION d ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g WASHINGTON. D. C. 20665 00*** June 12, 1998 The Honorable Shirley Ann Jackson Chairman U.S. NL. clear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
PROPOSED FINAL DRAFT OF THE NRC'S HUMAN PERFORMANCE PLAN During the 452nd and 453rd meetings of the Advisory Committee on Reactor Safeguards, April 30-May 2 and June 3-5,1998, respectively, we reviewed different drafts of the NRC's Human Performance Plan (HPP) Our Subcommittee on Human Factors reviewed an earlier draft of the Plan at a meeting on January 21,1998. During these reviews, we had the benefit of discussions with representatives of the NRC staff and of the documents referenced. The staff has informed us that it plans to rename the document as the "NRC Program to Assess Human Performance." As was originally intended, this document is an inventory of human performance activities within the NRC. It also defines a mission statement for the Human Performance Program, as follows: To ensure effective risk-informed and performance-based regulation and oversight of human performance in the design, operation, maintenance, and decommissioning of nuclear reactor sites and other NRC-regulated facilities by:
- 1) identifying human performance issues important to public health and safety; 2) increasing understanding of the causes and consequences of degraded human performance in such settings; and 3) implementing the appropriate regulatory response to such issues.
We agree with the mission statement. The document, however, does not describe a systematic cpproach for achieving the three goals of the mission statement. We continue to believe that a plan is needed to identify and prioritize activities related to human performance and to address the issues that we identified in our report dated February 13,1997, Cnd in our various communications to the staff. Sincerely, R. L. Seale Chairman I 79
References
- 1. Memorandum dated April 23,1998, from Thomas L. King, Office of Nuclear Regulatory Research, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
NRC's Human Performance Plan (Predecisional).
- 2. Memorandum dated May 29,1998, from Thomas L. King, Office of Nuclear Regulatory Research, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
NRC's Human Performance Plan (Predecisional).
- 3. Report dated February 13,1997, from R. L Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Human Performance Program Plan.
- 4. Letter dated October 8,1997, from R. L. Seale, Chairman, ACRS, to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
Human Performance and Human Reliability implementation Plan.
- 5. Letter dated December 30,1996, from T. S. Kress, Chairman, ACRS, to James M. 1 Taylor, Executive Director for Operations, NRC,
Subject:
ACRS Questions on Human Performance Program Plan. i l 1 80
r .-
/ UNITED STATES
/ o NUCLEAR REGULATORY COMMISSION j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS i o WASHINGTON, D. C. 20655 June 12,1998 l The Honorable Shiriey Ann Jackson Chairman U.S. Nuclear Regulatory Commission l
Washington, D.C. 20555-0001 l l Dear Chairman Jackson.
SUBJECT:
PROPOSED FINAL STANDARD REVIEW PLAN SECTION 3.9.8 AND
- REGULATORY GUIDE 1.178 FOR RISK-lNFORMED INSERVICE INSPECTION OF PlPING During the 451st,452nd, and 453rd meetings of the Advisory Committee on Reactor Safeguards, April 2-4, April 3M4ay 2, and June 3 5,1998, respectively, we met with representatives of the NRC staff, the Nudoar Energy Institute, Westinghouse Owners Group (WOG), Electric Power Research Institute, and others to discuss the proposed final Standard Review Plan Section 3.9.8 and essociated Regulatory Guide 1.178 (formerfy DG-1063) for risk-informed inservice inspection (ISI) of piping. We also had the benefit of the documents referenced. )
. Conclusions and Recommendations
- 1. We agree with the staff and the industry that a more risk-informed ISI program will result in reductions in the risk of piping failure, occupational radiation exposure to personnel, and -
associated inspection costs.
- 2. Regulatory Guide 1.178 should be edited prior to issuance to reduce redundancy with i Regulatory Guide 1.174.
- 3. Regulatory Guide 1.178 should be issued in final form rather than for " trial use." We agree l
with the staff's intent to document the technical content of the appendices from the draft Regulatory Guide in a draft NUREG and make it available to the public.
- 4. The staff should expedite and complete its review of the pilot applications and the Westinghouse topical reports as soon as possible.
Discussion l Current piping inspection programs are based on Section XI of the ASME Code. When the Code rzquirements were established, the degradation mechanisms that affect piping and methods for cssessing the risk significance of piping failures were less well understood. Both the staff and the industry believe, and we agree, that risk-informed ISI is expected to lead to reductions in the risk from piping failures, mWanal radiation exposure to personnel, and associated inspection costs. 81
2 Our review of the Regulatory Guide and the appendices that were part of DG-1063, as well as the presentations by industry, lead us to believe that these methods' can be used to develop risk-informed ISI programs. The Regulatory Guide provides general guidance for developing a risk-informed ISI program. The appendices to DG-1063 provide much more detail on a particular approach. We believe that the staff's decision not to include the appendices in Regulatory Guide 1.178 is correct. In general, regulatory gudes should focus on fundamental guidance while allowing the licensees flexhhty in demonstrating how they will apply such guidance. This should also permit the staff to focus its effort on cciiWng the safety evaluations for the topical reports and pilot plant applications. We also believe the appendices should be released as a NUREG quickly so that technical approaches will be available for use and comment, in many instances, Regulatory Guide 1.178 simply reiterates (or refers to) passages from Regulatory Guide 1.174 without evaluating their relevance and applicability to ISI. Examples include: the sections on defense-in-depth, safety margins, and acceptance criteria. Regulatory Guide 1.178 should be edited prior to issuance to reduce some of this redundancy. We believe that issuing Regulatory Guide 1.178 for trial use will send the wrong message to the industry concoming the staff's willingness to consider risk-informed ISI submittals. The Commission, in approving publication of Regulatory Guide 1.174 and the associated SRP Chapter 19 in the Federal Register, directed that"the staff should perform annual reviews of Regulatory Guide 1.174 and SRP Chapter 19 and incorporate experience gained from risk-informed pilot programs when revisions are necessary." Regulatory Guide 1.178 and SRP Section 3.9.8 could also be subjected to such annual reviews and future revisions. Consequently, release for trial use is unnecessary. The staff has informed us that it plans to complete its review of the pilot programs and the industry topcal repods by December 31,1998. Due to the importance of risk-informed ISI to the industry and the obvious benefits from its application in terms of reduced risk from piping failures, occupational radiation exposure, and associated inspection costs, we urge the staff to expedite and complete its l reviews as soon as possible. Dr. Dana Powers did not participate in the Committee's deliberation regarding Regulatory Guide 1.178. Sincerely, R.L.Seale Chairman
References:
- 1. Memorandum dated May 12, 1998, from Malcolm R. Knapp, Acting Director, Office of I Nuclear Regulatory Research, NRC, to Thomas T. Martin, Committee to Review Generic Requirements, NRC,
Subject:
Transmittal of Regulatory Guide 1.178: "An Approach for Plant-Specific, Risk-Informed Decision-Making: Inservice Inspection of Piping," and Standard Review Plan Sechon 3.9.8, " Standard Review Plan for the Review of Risk-Informed inservice inspection Applications."
- 2. Staff Requirements Memorandum dated May 21,1998, from John C. Hoyle, Secretary, NRC, to L Joseph Callan, Executive Director for Operations, NRC,
Subject:
SECY-98-015 - Final 82 l
3 General Regulatory Guide and Standard Review Plan for Risk-informed Regulation of Power
~
Reactors.
- 3. Westinghouse Energy Systems, WCAP-14572, Revision 1, " Westinghouse Owners Group AM#jm of Risk-Informed Methods to Piping inservice inspection Topical Report," October 1997.
- 4. Westinghouse Energy Systems, WCAP-14572, Revision 1, Supplement 1, " Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection," October 1997.
- 5. Report dated July 14,1997, from R. L Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Proposed Regulatory Guide and Standard Review Plan Chapter for Risk-informed, Performance-Based inservice inspection.
- 6. Letter dated June 2,1998, from Louis F. Liberatori, Jr., Westinghouse Owners Group, to U.
S. Nuclear Regulatory Commission, transmitting Additional Comments on the Draft Regulatory Guide and Standard Review Plan for Risk-informed Inservice inspection. i 83 l I
C*% g UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION
$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsHWGTON, D. C. 20555
***** June 15,1998 l
l The Honorable Shirley Ann Jackson l Chairman j U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 1 Dear Chairman Jackson.
SUBJECT:
NRC REACTOR FUELS RESEARCH PROGRAM ; l During the 453rd meeting of the Advisory Committee on Reactor Safeguards, June 3-5,1998, we discussed with the NRC staff the proposed plans for the NRC research program on reactor fuels. l Our Subcommittee on Reactor Fuels also discussed this program during a meeting on April 23-24, ) 1998. We also had the benefit of the document referenced. l BACKGROUND . There are large economic incentives for licensees to extend the bumup of reactor fuels. Extended l reactor fuel bumup also has .;yJ,csit societal benefits. It is not surprising, then, that over the last i two decades fuel bumups at discharge from reactor cores have more than doubled. Today, some nuclear units are approved to extend bumup to 62 GWd/t (peak rod average). The nuclear industry has indicated an interest in further extending fuel bumups to as high as 75 GWd/t. The nuclear , industry is currently reluctant to go to yet higher bumups, since fuel enrichments in excess of 5% l would be required. l In the past, the NRC maintained an active experimental research program to study the performance of reactor fuels under accident and off-normal conditions. The NRC developed codes (FRAPCON and FRAPTRAN) for predicting changes in fuel and fuel cladding with bumup and these codes are ! used in reviewing and approving licensee proposals for core reloads. This research program i' snabled the NRC to establish fuel performance criteria. Licensees were expected to perform analyses and tests needed to demonstrate that their fuel met these criteria. Funding reductions and the press of other needs forced the NRC to curtail its fuel performance research at a time when the experimental database extended to only about 33 GWd/t. Similarty, models of fuel and cladding ! properties were restricted to this limited database. Although the models were developed based on l a database for low bumup fuels, they would still be acceptable at higher bumups if there were no significant changes in the fuel or the cladding. Unfortunately, changes do occur in fuel and cladding starting at bumups in the vicinity of 50 to 60 GWd/t. The fuel develops a high porosity " rim" of low thermal conductivity. The cladding can undergo rapid or" breakaway" oxidation. Zirconium hydrides precipitate and embrittle the cladding. The effects of these changes in high bumup fuels have been demonstrated in French arad Japanese , tIsts of high bumup fuel behavior during reactivity insertions such as might be caused by design-basis control rod ejection acodents or control rod drop acadents. Cladding rupture and fuel dispersal were observed in the tests at energy inputs of 1/3 to 1/10 the levels expected based on current regulatory guides. Analyses of the test results show that cladding oxidation and embrittlement are 85
2 important contributors to the poor fuel performance.' These findings raise questions about the capacity of high bumup fuel to survive other design-basis accidents such as loss-of-coolant accidents (LOCAs) and anticipated transients without scram (ATWS). The test findings and operational events, such as control rod sticking and axial distortion of the neutron flux in reactor cores, show that high bumup fuel does not behave in ways anticipated by simple extrapolation of data forlower bumups. The problems are made more complicated by some evidence that the route to high bumup as well as the bumup level may affect fuel performance. Regulatory response to these findings has been to limit bumups to 62 GWd/t. The Office of Nuclear Reactor Regulabon (NRR) has concluded that the degradation of fuel and cladding at these bumups does not pose a 4 din.id threat to the public health and safety. The Office of Nuclear Regulatory Research (RES) has been asked to conduct research to confirm this regulatory decision. RES has not been asked to resume the broad program of exploratoiy research that it had in the past. Licensees will be expected to provide all the data and analyses needed to support approval to extend fuel bumups beyond the current limit. In addition, it is expected that licensees will implement more aggressive lead test assembly programs and will establish fuel performance monitoring programs. THE CONFIRMATORY RESEARCH PROGRAM RES has formulated a confirmatory research program that consists of the following elements:
. continued collaboration with intemational experimental studies of fuel behavior during reactivity insertions, . experimental studies of high bumup fuel behavior under LOCA conditions, e experimental studies of high bumup fuel behavior under ATWS conditions, . determination of the oxidation behavior and mechanical properties of cladding as a function of bumup, . upgrading properties correlations in the FRAPCON and FRAPTRAN codes, and . analyses of uncertainties in neutronic codes used by NRC.
The experimental and analytical studies in this program are limited to the confirmation of the regulatory decision to permit bumups to 62 GWd/t. RES has used risk insights to focus experiments and analyses on the issues of most importance. Resources have been leveraged by collaboration with continued intemational programs and with an industry-sponsored program. In addition to the RES confirmatory research program, the Office of Nuclear Material Safety and Safeguards (NMSS) proposes to benchmark its criticality models for fuel with enrichments in excess of 5%.
- COMMENTS AND RECOMMENDATIONS We make the following comments and recommendations:
. The strategy to require licensees to provide all the data and analyses to support extension 86
3 of fuel bumups beyond current limits seems appropriate. It places the burdens and responsibilities on those who will gain the rewards that come from extending fuel bumup. The strategy does, however, place limits on the technical independence that the NRC will have in establishog fuel performance acceptance criteria. NRR should make it clear that this strategy applies also to bumups within current limits for fuels with new cladding types not previously tested.
- The NRC should ensure that it has the knowledge and the tools to respond quickly to adequately formulated proposals from licensees to extend fuel bumups beyond 62 GWd/t.
"The staff needs to make clear what data and what analyses will be required to gain approval for extended fuel bumup.
. The overall RES confirmatory research program is well conceived and deserves the support of the Commission.
- The expenmental studies of fuel behavior under LOCA conditions need to be augmented to include tests with more realistic time-temperature histories that may impose harsher thermal and mechanical stresses on the cladding and cladding oxides to ensure that Appendix K (10 CFR Part 50) requirements are adequate for high bumup fuels. Greater realism in the LOCA tests may be especially important if, as is now expected, licensees take advantage of the option of using realistic analyses to comply with the requirements of Appendix K. Time-temperature histories need to be selected such that realistic or at least conservative thermal and mechanical stresses are placed on cladding oxides and any potential for breakaway oxidation is revealed. The suggestion from NRR that the route to bumup as well as the bumup level affects fuel performance needs to be addressed by the research program. Tests with more cladding types than are now approved for high bumup operation may be needed.
The test matnces would benefit from application of well-known experiment design methods.
- Plans to address issues of fuel performance during ATWS events have not yet been developed. ATWS may be an especially critical event for high bumup fuels.
- The conclusion that high bumup does not affect radionuclide source terms used in regulatory safety analyses is not supported by the technicalliterature. The research program should include consideration of how bumup may affect core degradation behavior under severe accidents. It may be necessary to confirm the validity of existing core degradation models used to estimate risk for high bumup fuels.
. The confirmatory research program needs to be augmented with an anticipatory component to give the NRC line organizations the tools to respond to inevitable proposals from licensees for extended fuel bumup. RES has stated that fuel performance models are being refurbished to better predict the thermal and mechanical loads on cladding, as well as the embrittlement of the cladding. We have heard no details on this portion of the program. Furthermore, RES should begin now to develop criteria for use by the licensees in proposing fuel performance monitoring programs. . We can find no immediate justification for work proposed by NMSS for fuel with enrichments in excess of 5%. The nuclear industry indicates a reluctance to use fuels that have higher ennchment. Should this change, there will be ample time to carry out activities proposed by NMSS. 1 87
4 Dr. William Shack did not participate in the Committee's deliberation regarding this matter. Sincerely, l R. L Seale Chairman
Reference:
Memorandum dated May 22,1998, from Thomas L King, RES, to John T. Larkins, Executive Director, ACRS,
Subject:
Transmittal of Advance Copy of Agency Program Plan for High-Bumup Fuel (Predecisional) I 4 88
ag i 8 g\ UNITED STATES { o NUCLEAR REGULATORY COMMISSION i
$ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l wAssincrow, p. c.aoses e....
June 15,1998 Mr. L Joseph Callan Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555 0001
Dear Mr. Callan:
SUBJECT:
THE SAFETY ASPECTS OF THE WESTINGHOUSE ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE AP600 PLANT DESIGN - INTERIM LETTER 3 During the 453rd meeting of the Advisory Committee on Reactor Safeguards, June 3-5,1998, we i reviewed the AP600 test and analysis program, various chapters of the AP600 Standard Safety Analysis Report (SSAR), the Level 2 and 3 AP600 Probabilistic Risk Assessments (PRAs), severe accidents, regulatory treatment of non-safety systems, and the associated chapters of the NRC staft's advance Final Safety Evaluation Report (FSER). Our Subcommittees on Thermal Hydraulic
, Phenomena and Advanced Reactor Designs reviewed these items on May 11-12 and May 13-15, 1998, respectively. During these reviews, we had the benefit of discussions with representatives of the NRC staff and the Westinghouse Electric Company. We also had the benefit of the documents referenced.
Based on our review to date, no additional issues were identified that would prevent the certification of the AP600 design. Our assessment is based, in part, on agreement by Westinghouse to improve its documentation of the test and analysis program. In addition, we identified several issues related to NRC staff assessment of accident phenomena. Our comments are provided below. TEST AND ANALYSIS PROGRAM in our interim letter dated February 19,1998, we identified a list of outstanding thermal-hydraulic issues related to the documentation of the reactor coolant system and containment designs. The issues related to the containment were discussed by our Thermal Hydraulic Phenomena Subcommittee on June 11-12,1998. Westinghouse responded to the issues related to the reactor coolant system at the May 11-12,1998 Thermal Hydraulic Phenomena Subcommittee meeting, and committed to perform additional analyses and studies, and to provide additional explanations. Based on our assessment of the Westinghouse responses, we are satisfied that Westinghouse has fulfilled several commitments by:
. Performing a sample analysis of the small-break loss-of-coolant accident (LOCA) involving automatic depressurization system (ADS) activation through initiation of in-containment refueling water storage tank (IRWST) flow to show the relationship between the IRWST level 89 1 l
l m..._ - - - - - - i
. penalty and flow resistances in the ADS piping. This analysis provided assurance that the level penalty Westinghouse takes in the NOTRUMP small-break LOCA code is an appropriate and conservative compensation for neglecting the momentum flux terms in the blowdown equation.
- Amplifying the Westinghouse scaling analysis to include the reistionships between core inventory and the multiple flow paths This permitted evaluation of the usefulness of relevant data obtained from the Oregon State Uruversity and the SPES-2 test facilities during the ADS l actuatum phase of an acculent.
. Explaining the difference in timing for the minimum reactor vessel water level between the !
value calculated with the NOTRUMP code and the test data. ! Westinghouse still needs to submit the following additional information:
. The results of the break area sensitivity study for one of the severe small-break LOCAs to ensure that the process for compensating for exclusion of momentum flux terms in the NOTRUMP code is robust for a range of blowdown rates, t . A discussion of the implications of the sensitivity of the results to the assumed heat loss distnbubon in the SPES-2 test facility in validating both the LOFTRAN and NOTRUMP codes.
A description in SSAR Chapters 4 and 15 of the interrelationships among the LOFTRAN, j THINC-IV, and WESTAR codes and the test data. Clear identification of channel-to-channel mixing coefficients to be used. Clearidentification in the SSAR of the inadequacies in the NOTRUMP code and the steps taken to compensate for them. ST_MRO SAFETY ANALYSIS REPORT AND THE ADVANCED FINAL SAFETY EVALUATION REPORT We reviewed the Level 2 and 3 PRAs, severe accidents, regulatory treatment of non-safety systems, the following SSAR chapters, and the associated NRC staffs advanced FSER:
. Chapter 3 - Design of Structures, Components, Equipment, and System . Chapter 6 - Engineered Safety Features . Chapter 14 - Initial Test Program . Chapter 16 - Technical Specifications . Chapter 17 - Quality Assurance Based on our review of the above, we offer the following comments:
Reaulatory Treatment of Non-Safety Systems The active systems in the APS00 are designated as non-safety whereas, in existing plants many of these active systems are designated as safety related. The regulatory treatment of these non-safety systems, which are relied upon for defense-in-depth and to meet plant investment protection goals, is an excellent example of a good risk-informed and ! performance-based regulatory approach.
)
90
Chapter 9 - A*w Systems. InMina Aaaandix 9A - Fire PidMn Ar.fsis Smos issuing our second interim letter concoming the AP600 plant design on Apnl g,1998, i we have completed our review of the fire protection system design and the fire protection ! analysis The NRC staff has agreed with the Westinghouse proposal that the AP600 design i should be govemed by 10 CFR 50.48, " Fire protection." The AP600 fire protection analysis I used the Fire induced Vulnerability Evaluation (FIVE) screening methodology. The NRC staff review of this analysis identified that the original design did not provide separate water supplies for the fire fighting capabilities. Although Westinghouse did not agree that an additional water supply was needed, Westinghouse modified the design by relocating the diesel-driven fire pump from the turbine building to a prefabricated enclosure to be located in the yard. This proposed modification by Westinghouse will provide a separate water supply for fire fighting. Such a modifics.Mii brings the AP600 design into compliance with the requirements of 10 1 CFR 50.48 and the enhanced fire protection criteria approved by the Commission. Consequently, we conclude that the AP600 fire protection system design is adequate. l Environmental Qualifiemfian Tests for Pas *ive Autor=*=8vtic R=ceTibiriers l Supported platinum or palladium catalysts will be used to control hydrogen concentrations in the AP600 reactor containment following design-basis accidents. Such catalysts are i known to be fully capable of providing hydrogen recombination sufficient to meet regulatory 1 and safety requirements. Catalytic recombiners are susceptible, however, to deactivation during protracted use due to: ! e poisoning of the catalytic surface, e coking that occludes catalytic surfaces,- surface diffusion and sintering of catalytic materials that result in a loss of active surface area, and e interactions of noble metal with the substrate, j The effect of these processes is cumulative as the time of recombiner operation increases. Some short-term tests have examined the susceptibility of hydrogen recombiners to poisons ' and coking. Some of these tests are of questionable utility. The tests first exposed the catalysts to the poisoning material and then, in separate tests, measured the capacities of the . exposed catalysts to recombine hydrogen. Any synergistic effects of poison and recombination activity would not have been revealed by this procedure. Similariy, effects of radiolytically generated ozone and nitrous oxides were not examined in the tests. On the other hand, the tests have examined a wide range of materials that might be expected to adversely affect catalytic activity and only modest (<20%) reductions in catalytic activity were found in the short-term tests. Tests that simultaneously examine prototypic environments of temperature, radiation field, and catalytic activity for appropriate service times do not seem to have been done. The adverse effects shown in short-term tests may well become more significant as service , continues. Synergistic effects of radiation may exacerbate effects that are small under thermal conditions. 91
_4_ To increase confidence that the passive autocatalytic hydrogen recombiners will perform their intended functions effectively, there is a need for better environmental qualification tests. This may well be the responsibility of the Combined License (COL) applicant if, for no other reason, catalysts can be expected to be improved between now and the time a license is sought to operate an AP600 plant. We recommend that environmental qualification tests for passive autocatalytic recombiners include requirements for timing of exposure and exposure to pyrolysis products. ITEMS FOR CONTINUED STAFF EVALUATION FOR LICENSING ACTIONS Although the following items have been adequately addressed for the AP600 design, additional cvaluation of these items is needed to support efficient review of future license applications or licensing schons: ' Leak-Before-Break Evaluation of Feedwater Pipina The leak-before-break (LBB) criteria require that piping have high fracture toughness and not suffer from modes of degradation such as flow-assisted corrosion or stress-corrosion cracking that could result in significant loss of strength before detectable leakage occurs. The peing must also not be subject to large loads that were not accounted for in the original design, such as those which might result from a large water hammer. The NRC has developed guidelines and procedures (NUREG-1061, Vol. 3) that can be used to demonstrate that piping will exhibit LBB behavior. The AP600 design makes more extensive use of the LBB concept in the design of reactor system piping than current reactors. In the advanced FSER, the staff has concluded that Westinghouse has been able to demonstrate through the chok:e of materials for the piping, stress and fracture mechanics analysis procedures, and the controls placed on water chemistry, that the piping for safety-related systems meets the LBB guidance in NUREG-1061, Vol. 3. The staff denied the request to apply the LBB concept to the feedwater piping design. The staff agrees that the present design meets all the LBB guidelines in NUREG-1061, Vol. 3, except for susceptibility to water hammer. The staff also agrees that the feedwater piping i and steam generator designs for the AP600 have incorporated the " lessons leamed" from I operating plants for avoiding water hammers and that the piping design meets all the design guidelines for reducog susceptibility to water hammer. The staff argues, however, that there is no operating experience applicable to the AP600 design to demonstrate that the probability of a large water hammer it sufficiently low, and proposes a bounding water hammer load that , is 10 times as large as that proposed by Westinghouse. Since Westinghouse determined l that it was impractical to design the piping for a pressure pulse this large, Westinghouse l agreed to drop the request to apply the LBB concept to the feedwater piping. The bounding water hammer load proposed by the staff is based on the assumption that the main feedwater line fills with steam and then a large slug of cold water at high velocity is introduced into the pipog The staff concedes that the sequences of events that might lead to such a water hammer would require misalignment of several valves, but did not attempt to estimate the probability of such an event. According to Westinghouse, in order to establish the initial conditions assumed by the staff, the steam generator water level would have to be at a point that would trip the reactor. All procedures for refiilng a steam generator following 92
a reactor trip require using the auxiliary feedwater system, which injects water through a separate auxiliary feedwater injechon line. The bounding water hammer based on injection of cold water into the auxiliary feedwater line results in much smaller loads than those calculated by the staff for the main feedwater line. We believe that the staff should reexamine its position on the likelihood of the initial condibons assumed in calculatmg the load used in its bounding analysis for water hammers in feedwater piping. The staff has stated that it feels that some operational experience should be obtained with the AP600 feedwater system before approving the application of the LBB concept to the feedwater piping. It is completely impractical to demonstrate by operatonal monitoring the degree of assurance against large water hammers sought, which is <10* events / year. The degree of assurance could, however, be demonstrated by PRA 1 techniques, which could be benchmarked by comparing the results of such analyses for current feedwater piping systems with operational experience. In-Vessel Retention An AP600 strategy for mitigating the consequences of severe accidents is in-vessel retention of molten debris through extemal cooling of the reactor vessel. The reactor cavity is flooded with water to provide cooling of the lower head. A substantial experimental program using scaled models and sections of the lower head to support the heat transfer analyses has been used to evaluate the retention of the core melt. These tests, however, have not used prototypic materials. The analysis of in-vessel retention performed for the AP600 fails to demonstrate convincingly that vessel failure during a core melt is extremely unlikely. This analysis relies on a specified melt geometry in the lower head and considers only decay heat and stored energy. The possibility of a zirconium-iron exothermic interaction leading to vessel failure has not been adequately considered The existence of such intermetallic exothermic reactions could alter the severe accident picture for future analyses and should be further investigated. The models and analyses used to develop this core degradation scenario have not been validated against experiments involving large volumes of molten metals and molten oxides. The deficiencies of the core degradation modeling afflict both the likelihood of in-vessel retention of core debris and the susceptibility of the reactor to in-vessel steam explosions. The RASPLAV experimental activities supported by the NRC are not likely to resolve the most i,T,ponent issues of materialinteractions involving in-vessel retention. Results of these experiments will not be useful in studying the effects of mixing a large volume of molten metal with hypostoichiometric reactor fuel. Based on the results of the analysis, Westinghouse concludes that it is " physically unreasonable" for the vessel to be penetrated by molten core debris. The NRC staff, on the other hand, has concluded that the possibility of reactor vessel penetration cannot be excluded. We agree with the staff's conclusion. Even discounting retention within the vessel and assuming containment vulnerability, the AP600 poses low risks to the public relative to existing reactors because the AP600 has quite a low core damage frequency and because the cavity will be flooded. 93
Since in-vessel retention is widely considered to be an important accident management strategy for operateg reactors, the impact of intermetallic exothermic reactions on this i strategy should be assessed by the staff. CONCLUSION ! As noted above, we have identified no additionalissues that would prevent the certification of the AP600 design. We plan to complete our review of the AP600 design, including resolution of our previous concoms, at the July 1998 meetmg We contmue to be concemed about the quality of test and analysis program documentaten related to informaten needed to certify the AP600 design. The staff should evaluate whether the quality of the AP600 documentation could withstand an NRC design-basis inspection. Sincerely, l kb l R. L Seale Chairman References-l 1. Letter dated April 9,1998, from R. L Seale, Chairman, ACRS, to L Joseph. Callan, Executive Director for Operations, NRC,
Subject:
The Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design-interim Letter 2. l 2. Letter dated February 19,1998, from R. L Seale, Chairman, ACRS, to L. Joseph Callan, l Executive Director for Operatons, NRC, Subpect interim Letter on the Safety Aspects of the l Westinghouse Electric Company Application for Certification of the AP600 Plant Design. ! 3. U. S. Department of Energy Report DE-AC03-90SF18495 dated June 26,1992, prepared by l Westinghouse Electric Corporation "AP600 Standard Safety Analysis Report," updated I through Revision 22 (issued April 6,1998).
- 4. U. S. Department of Energy report DE-ACO3-90SF18495 dated June 26,1992, prepared by Westinghouse Electric Company, "AP600 Probabilistic Risk Assessment," updated l through Revision 11 (issued March 1998).
l S. U.S. Nuclear Regulatory Commission, " Advance Final Safety Evaluation Report Related to ' the Certification of the AP600 Design," dated May 1998 (Predecisional Information).
- 6. Westinghouse Electric Company, WCAP-14807, Revision 4, dated February 27, 1998, "NOTRUMP Final Validation Report for AP600" (Proprietary).
- 7. Memorandum dated March 13, 1998, from Brian A. McIntyre, Westinghouse Electric
, Corporation, to U. S. Nuclear Regulatory Commission, transmitting errate p::ges to WCAP-l 14807, NOTRUMP Final Validation Report for AP600, Revision 4.
- 8. Set of page changes to WCAP-14727: " Scaling and PlRT Closure Report", Volumes 1 and 2, to update the report to Revision 2.
- 9. Memorandum dated March 2,1998, from Brian A. McIntyre, Westinghouse Electric Corporation, to U. S. Nuclear Regulatory Commission, transmitting additional material for incorporation into WCAP-14727, AP600 PXS Scaling and PIRT Closure Report, Revision 2, RAI Responses for Appendix A. i
- 10. Westinghouse Electric Company, WCAP-14305, Revision 2, dated April 7,1998, "AP600 Test Program ADS Phase B1 Test Analysis Report"(Proprietary).
94
1 i j
- 11. Westinghouse Electnc Company, WCAP-14171, Revision 2, dated March 1998, "WCOBRA/ TRAC Applicability to AP600 Large-Break Loss-of-Coolant Accident" (Propnetary).
- 12. Memorandum dated April 20, 1998, from Brian A. McIntyre, Westinghouse Electric Corporation, to Robert Seale, Chairman, ACRS, transmitting "Roadmap* of Westinghouse Responses to ACRS Concems.
- 13. Letter dated April 28,1998, from Westinghouse Electric Corporation, to U. S. Nuclear Regulatory CviT.rhi,
Subject:
Revised Response to FSER Open item 440.796F, Part E, on the NOTRUMP Final Validation Report. j
- 14. Westmghouse Electnc Company, WCAP-14305, Revision 3, dated April 1998, "AP600 Test l Program, ADS Phase B1 Test Analysis Report"(Proprietary).
- 15. . Westinghouse Electric Company, WCAP-14776, Revision 4, dated March 1998, "WCOBRA/ TRAC OSU Long-Term Cooling Final Validation Report" (Proprietary).
- 16. Letter dated April 9,1998, from T. H. Essig, Office of Nuclear Reactor Regulation, NRC, to N. J. Upa uio, Westinghouse Electric CGTs.Gw,
Subject:
Documentation of Topical Report WCAP-12945(P) " Westinghouse Code Qualification Document for Best Estimate Loss of Coolant Analysis." 95
/ 'o g UNITED STATES
/ o NUCLEAR REGULATORY COMMISSION fo ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 o**** June 16,1998 The Honorable Shirley Ann Jackson Chairman U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
REVIEW OF SECY-98-076," CORE RESEARCH CAPABILITIES" During the 453rd meeting of the Advisory Committee on Reactor Safeguards, June 3-5,1998, we reviewed the subject document. Our Subcommittee on Safety Research Program met on June 1, 1998, to review this matter. During these reviews, we had the benefit of discussions with representatives of the NRC staff. We also had the benefit of the documents referenced. CONCLUSIONS
- 1) The twenty-nine areas of core capabilities identified in SECY-98-076 are clearly too many to be supported, either philosophically or considering budget constraints.
- 2) There is a need for a better definition of core research capabilities which incorporates the dimension of "essentiality" for NRC to effectively cany out its mission.
BACKGROUND SECY-97-075," Methodology and Criteria for Evaluating Core Research Capabilities," was prepared by the Office of. Nuclear Regulatory Research (RES) in response to the Commission's Direction Setting issue (DSI) 22. "Research." This, along with other guidance, stated that "the NRC will maintain a core research capability now and in the future to support NRC's regulatory function." The Comrnission approved SECY-97-075 and provided additional guidance which resulted in an intensive, year-long review of the RES programs and the development of a systematic process described in SECY-98-076 for the identification of " expertise driven" core research capabilities. This required that the staff make an evaluation of the expertise deemed to be vital to NRC's ability to regulate nuclear facilities and programs. In response to this requirement, the staff formulated a list of existing core competencies and then used an extensive but subjective evaluation process to rate each candidate competency. Twenty-nine areas of core research capabilities were identified by this process, and associated core levels of resources within the NRC and its contractors were specified. The expertise-driven programs are those areas of technology deemed essential to the long-term (over several years) effectiveness of the regulatory process. Facilities (e g., hot cells capable of handling examinations of full-length spent fuel assemblies) and expertise (e.g., nuclear materials technologies) are to be maintained regardless of immediate need. Work under these programs would also involve anticipatory research to address issues that NRC expects to face in the future. 1 97
l l DISCUSSION The designation of" core research capabiktes" essental for the NRC to effectively fulfill its regulatory mission is an important exercise that could have the ancillary benefit of providing the rationale and I justification for maintaining a viable and robust research component within NRC. Given the importance of doing this, we feel that the effort (and the SECY-96-076 results) falls short of providing a useful departure point for achieving this desirable objective. The staff concluded that there are twenty-nine areas of core capabilities. This conclusion was neither supported by the information provided nor can it be justified based on the budgetary levels. What is needed is a better definition of core research capabilities which incorporates a dimension .of *essentiality" for NRC to effectively carry out its mission. " Effectively" here implies ensuring acceptable risk, providing timely response to incidents and emerging issues, and controlling ( excessive burden on the industry. The identification of such core capabilities must involve an awareness of the uniqueness of the informational needs associated with safe and efficient utilization of inuclear technology. Moreover, the " selection criteria" used for selecting among the range of I candidate competencies must provide a clear discrimination based on the elements of risk and benefit addressed in each candidate competency. The appropriate process should be to identify the activites required to meet the NRC mission, and then select only the associated capabilities that are unique in their application to nuclear technology or for which independence in technical assessment is essential. Although prioritization was not part of the Commission's request, we believe that the evaluation process should provide a basis for discriminating among research areas within the core research capabilities. Differentation among the selected core research c.f,et,iidies with respect to importance is essential when priontization of resources is required in the budgetary process. Dr. William Shack did not participate in the Committee's deliberation regarding this matter. Sincerely,
. --x R. L. Seale Chairman
References:
- 1. SECY-98 076, Memorandum dated April 9,1998, from L Joseph Callan, Executive Director for Operations, NRC, for the Commissioners,
Subject:
Core Research Capabilities.
- 2. SECY-97075, Memorandum dated April 2,1997, from L. Joseph Callan, Executive Director for Operations, NRC, for the Commissioners,
Subject:
Methodology and Criteria for Evaluating Core Research Capabilities.
- 3. Staff Requirements Memorandum dated June 6,1997, from John C. Hoyle, Secretary, to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
SECY-97-075 - Methodology and Criteria for Evaluating Core Research Capabilities. 98
# UNITED STATES
_f %,, NUCLEAR REGULATORY COMMISSION
'f ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS wAsmworow, p. c. 2oses July 16,1998 f
l l l I The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Cheirman Jackson:
l
SUBJECT:
PROPOSED REVISIONS TO 10 CFR 50.59 (CHANGES, TESTS AND EXPERIMENTS) During the 453rd and 454th meetings of the Advisory Committee on Reactor Safeguards, June 3-5 and July 8-10,1998, we met with representativesof the NRC staff and the Nuclear Energy institute (NEI) to discuss proposed revisions to 10 CFR 50.59 (Changes, Tests and Experiments). We also discussed the Executive Director for Operation's (EDO's) response to the Commission's directions in the Staff Requirements Memorandum (SRM) dated March 24,1998, regarding SECY-97-205. Our Subcommittee on Plant Operations met on June 19,1998, to discuss these matters. We also had the benefit of the documents referenced. We previously provided reports to the Commission l cn April 8, October 9, and December 12,1997, on the proposed revisions to 10 CFR 50.59 and related matters. During our meeting with the staff on July 8,1998, it became apparent that the staff had developed r; vised documents for consideration by the Commission. These documents were provided to us subsequent to our discussions with the staff. We have not had an opportunity to review these documentsin detail. In our preliminary review, however, we found substantial changes had been made to these documents. Conclusens and Recommendations
- 1. We disagree with the staff's evaluation of the proposed rulemaking language in response to the SRM dated March 24,1998. Therefore, we recommend that the revised rulemaking package not be issued for public comment at this time.
- 2. We believe that the revised guidance is overly prescriptive in defir.ing reduction in margin of safety. The revised rulemaking language will likely add significant regulatory burden without a clear safety benefit.
99
2
- 3. The staff's proposed use of ANSI /ANS-58.8-1994, " Time Response Design Criteria for SafetyRelated Operator Actions," for determining when there is only a minimal increase in the probability of malfunction is inappropriate.
- 4. The staff should expedite completion of a regulatory guide for implementing 10 CFR 50.59, including endorsement of NEl 96-07 with exceptions and clarifications as appropriate.
Docussion We disagree with the staff's evaluation of the rulemaking language provided in the EDO's response to the SRM dated March 24,1998. We believe that the staff should complete the actions specified by the Commission in the SRM. The staff's approach to reductions in margin of safety is overly prescriptive. We believe that the language proposed by the staff creates a de facto constraint of "zero increase" in probability or consequences. In addition, the lack of a working definition of " minimal" increase in risk may exacerbate the problem of excessive resources being required to perform evaluations for changes that are risk insignificant. In our November 14,1995 report, we informed the EDO that we found no technical basis for the estimates of minimum times for operator actions specified in ANSI /ANS-58.8-1994. We did not support the staff's proposed endorsement of ANSI /ANS-58.8-1994in the proposed final Regulatory Guice 1.164 and stated that we did not believe that this endorsement was the appropriate way to resolve Generic Safety lasue B-17, " Criteria for Safety-Related Operator Actions." Subsequent to the meeting, the staff informed us that it was unaware of our position and agreed to delete ANSI /ANS-58.8-1994 from the revised rulemaking package. The staff has made some progress in reconciling its differences with NEl 96-07. However, more than a year has passed and the staff has not codified its positions through development of a regulatory guide. We believe that development of a regulatory guide and endorsement of appropriateindustry guidance are essentialto stabilize the 10 CFR 50.59 process. Therefore, we believe that a regulatory guide should accompany the proposed rulemaking and associated documents in soliciting public comment. In our October 9,1997 report, we encouraged the continued development of a plan for a 10 CFR 50.59 process that is consistent with risk-informed, performance-based regulation. In our December 12,1997 report, we stated that the development of a risk-informed rule should be continued on an expeditious schedule. We continue to believe that 10 CFR 50.59 can l accommodate risk-informed decisionmaking. We believe that a deterministic regulatory framework poses substantial barriers to the developmert of the concept of minimal changes in accident probabilities or consequences. It is more likely that the minimal change concept can be defined satisfactorily within the framework of frequency-consequence (F-C) curves. A frequency-consequence framework can also accommodate the evaluation of new accident sequences. Dr. Apostolakis has offered a proposal (attached) for the development of a risk-informed framework for 10 CFR 50.59. We are examining how such a 100 l
s 3 l framework might be incorporated in developing an improved 10 CFR 50.59 process. We plan to address this in a future report to the Commission. I Sincerely,
/f 1. J R. L. Seale l
Chairman
References:
- 1. Draft Commission paper from L. Joseph Callan, Executive Director for Operations, NRC, to the Commissioners,
Subject:
Proposed Rulemaking on 10 CFR Parts 50,52 and 72 RequirementsConceming Changes, Tests and Experiments and Staff Recommendations on Changes to Other Regulations and Enforcement Policy, and attachments, received July l 8,1998.
- 2. Memorandum dated May 21,1998, from David B. Matthews, Office of Nuclear Reactor l Regulation, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
Proposed l Rulemaking on 10 CFR 50.59. l 3. Memorandum dated May 27, 1998, from L. Joseph Callan, Executive Director for l Operations, NRC, to the Commissioners,
Subject:
Evaluation of Rulemaking Language Proposals Conceming 10 CFR 50.59 (Changes, Tests and Experiments).
- 4. Memorandum dated March 24,1998, from John C. Hoyle, Secretary of the Commission, l to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
Staff Requirements l SECY-97-205, Integration and Evaluation of Results from Recent Lessons-Leamed l Reviews. , 5. Letter dated May 12,1998, from Samuel J. Collins, Office of Nuclear Reactor Regulation, 1 NRC, to Ralph E. Beedle, Nuclear Energy institute,
Subject:
10 CFR 50.59 Safety l Evaluations and 10 CFR 50.71(e) Final Safety Analysis Report Updates.
- 6. Letter dated April 16,1998, from Ralph E. Beedle, Nuclear Energy Institute, to Shirley Ann Jackson, Chairman, NRC,
Subject:
10 CFR 50.59 Safety Evaluations and 10 CFR 50.71(e) FSAR Updates. ! 7. Letter dated January 9,1998, from Samuel J. Collins, Office of Nuclear Reactor Regulation, l NRC, to Ralph E. Beedle, Nuclear Energy Institute,
Subject:
NEl 96-07 Guidelines for 10 CFR 50.59 Safety Evaluations.
- 8. Report dated December 12,1997, from R. L. Seale, Chairman, ACRS, to Shirley Ann l Jackson, Chairman, NRC,
Subject:
Proposed Revisions to 10 CFR 50.59 (Changes, Tests I and Experiments).
- 9. Report dated October 9,1997, from R. L. Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Proposed Changes to 10 CFR 50.59 and Proposed Revision 1 to Generic Letter 91-18.
- 10. Report dated April 8,1997, from R. L. Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Proposed Regulatory Guidance Related to implementation of 10 CFR 50.59 (Changes, Tests and Experiments). !
- 11. Report dated November 14,1995, from T. S. Kress, Chairman, ACRS, to James M.
Taylor, Executive Director for Operations, NRC,
Subject:
Proposed Final Regulatory Guide 101 l 1
4 1.164, " Time Response Design Criteria for Safety-Related Operator Actions," to Resolve Generic Safety issue B-17.
- 12. U.S. Nuclear Regulatory Commission, NUREG-0933, Supplement, March 16,1987,"A Prioritization of Generic Safety issues," item B-17, " Criteria for Safety-Related Operator Actions," Revision 2.
Attachment:
A Proposalforthe Development of a Risk-Informed Framework for 10 CFR 50.59 and Related Matters Prepared by ACRS Member Dr. George Apostolakis 102
1 ATTACHMENT A PROPOSAL FOR THE DEVELOPMENT OF A RISK-INFORMED FRAMEWORK FOR 10 CFR 50.59 AND RELATED MATTERS Prepared by ACRS Member Dr. George Apostolakis Preamble While I agree with my colleagues on what needs to be done in the near term, I believe that the issues that the revision of 10 CFR 50.59 has raised stem from some fundamental problems that cannot be eliminated by revising this particular regulation in isolation. A bolder approach is required. 10 CFR 50.59 is intended to limit to below a " minimal" level the impact of unreviewed changes in the plant on the probability or consequences of accidents. The point of reference is clearly the status of the plant before the change, as it has been approved by the staff. This approvalis based on traditional" deterministic" calculations, as documented in the plant's Final Safety Analysis Report (FSAR). The staff's interpretationis that the changes ought to be minimal compared to the existing situation and not with respect to regulatory criteria such as a 25 rem exposure to the whole body cnd a 300 rem exposure to the thyroid. T"a Nuclear Energy Institute disagrees and argues that the point of reference ought to be the reou. story criteria. In other words, changes that are within the operating margin should not be subjected to staff review. I understand that, in the majority of cases,10 CFR 50.59 works very well and there is no controversy. There are several changes per plant per year, however, in which there are questions es to whether the changes are minimal and, consequently, whether staff review is required. Direction from the Commission in the Staff Requirements Memorandum dated March 24,1998, the Commission directed the staff 'to incorporate knowledge gained from risk insights, as appropriate" and to " consider the work it has done in updating the Standard Review Plan (NUREG-0800) and, in a different context (severe Eccusents), on draft Regulatory Guide 1.174 and Standard Review Plan Chapter 19 in formulating what constitutes a ' minimal' change." The Commission also directed the staff to " evaluate for Commission consideration the advisability of allowing proposed changes that result in the creation of an accident or malfunction of a different type than previously evaluated that has ' minimal' safety impact." The Problem A major difficultyis defining the concept of a " minimal" increase in the probability of an " accident" and of a " minimal" change in the probability of equipment malfunction (which is treated separately from the probability of an " accident"). 103
2 in the memorandum dated May 27,1998, to the Commission, the EDO stated that any plant change that created the possibility of a new accident that had not been previously evaluated should be reviewed. The EDO also expressed the opinion that "it would be extremely difficult to develop a meaningful definition of minimal safety imjact." What are the Root Causes of the Problem? Concepts such as a " minimal" change in probability or consequences and " minimal" safety impact are meaningful only in the context within which the evaluation is made. SometN, g th&t i: minimal in one context may be significant in a different context. The deterministic regulatory context is incomplete for such evaluations. Whatis the Proper Context? l The proper context for evaluating changes in the spirit of 10 CFR 50.59 should include the information contained in the frequency-consequence curve.s (F-C curves), as well as the sets of accident sequences that probabilistic risk assessments (PF:As) produce. Figure 1 is an example of such F-C curves developed in the NUREG-1150 studies. The independent variable is the release fraction of the core inventory of iodine. The ordinate gives the frequency of the release exceeding a given value, R*. As an example, we see in Figure 1 that the frequency of sequences leading to a release fraction greater than 104 is, for Serry, about 7x104 per reactor-year. I note that these curves deal with accidents that involve the core only. These curves are an important element of the context within which we can evaluate what is minimal. Let us assume that a proposed change affects an accident sequence whose consequence is a release fraction greaterthan 104 and whose mean frequency is on the order of 104 per reactor-year, it is evident that changes that even double the frequency of this accident sequence could be considered minimal. If, on the other hand, the frequency of that accident . sequencewere on the orderof 104 per reactor-year, such a change would not be minimal (in the sense that it would have to be rev'ewed). If we consider release fractions greater than 104, then the average frequency of the accident sequences leading into that interval is about 2x104 per reactor-year (for Surry), i.e., about a factor of three greater than the frequency of the interval Rt104 Obviously, a " minimal" change in frequency would have a different numerical value for accidents in this new interval. l I note that these determinations are made with regard to the whole sequence and not separately for" accidents"and " equipment malfunctions,"as the current 10 CFR 50.59 requires. The latter is ( unnecessarily intrusive. I The same reasoning can be applied to "new" accident sequences. What would determine whether they had " minimal" impact would be their frequency of occurrence and their consequences. Thus, if the consequences placed the new sequence in the release internal R2104, then " minimal" would be defined with respect to the reference value of 7x104 per reactor-year, as just described. If, on the other hand, the consequences were in the interval Rt10", then the reference value for the frequency would be 2x104 per reactor-year. 104
i l 3 I hasten to add that the preceding paragraphs are not intended to imply that final decisions would have to be made solely on the basis of the F-C curves. The process of " integrated decisionmaking that is described in Regulatory Guide 1.174 would also apply here (with appropriate modifications) in the same spirit as that of Regulatory Guide 1.174, I am proposing to expand and modify the j decisionmaking process of 10 CFR 50.59 to include risk information. Doing so would definitely be ' responsive to the Commission's direction that the work done on Regulatory Guide 1.174 be considered in revising 10 CFR 50.59. This simple approach would obviate unnecessary debates regarding the operating and safety margins of the facility and how they ought to be handled. Are Quantitative Assessments Always Required? No. I believe that most cases could be handled qualitatively within this context. In other words, the accident sequences and their frequencies would add significantly to the basis for assessing qualitatively whether a change is minimal. I anticipate that a quantitative assessment will be required in very few cases. Is the Necessary Information Available? The F-C curves for severe accidents are either available or can be obtained from existing PRAs or Individual Plant Examinations (IPEs). These would have to be supplemented by the results for accidents that the regulations call Classes 1 through 8. This should not be a major problem, however. A first attempt to develop such curves for Class 3-8 accidents was made about 20 years ago in NUREG/CR-0603(Reference 1). It should be a straightforward process to produce similar results with modem PRA tools and using our current state of knowledge. The incompleteness of the PRAs/IPEs would be handled in a manner similar to that described in Regulatory Guide 1.174. Can the Regulatory Guide 1.174 Process be Extended to the Regulation of Lower-Class i Accidents? 1 Yes. What I discussed above dealt with changes that would not require staff review. The intent l of Regulatory Guide 1.174 was, of course, to define changes that, after review, would be acceptable. The same idea can be applied to the whole spectrum of aed.w.cs. Severe accidents are the contributors to releases of about 10% or greater of the iodine inventory (Figure 1). Regulatory Guide 1.174 provides acceptable ranges for changes in the frequency of a subset of these releases, namely releases that are "eariy," that is, those that are caused by accidents in which the containmentis either bypassed or fails before vessel breach (these control the prompt fatalities). Gqure 2 shows the guidelines adopted in Regulatory Guide 1.174 for large, early release frequer e, (LERF). To apply the Regulatory Guide 1.174 approach to releases other than those caused by severe 4 accidents,we must define goals for the frequency of smaller releases similar to the goal of 10 per reactor-year for large, early releases (Figure 2). 105
4 A way of defining goals for the frequency of lower releases is by defining appropriate Farmer curves. An example is given in Figure 3. The issue is what should the position of this curve be to ensure consistency with the Commission's stated Quantitative Health Objectives and subsidiary objectives (the region of severe accidentsis indicatedin this figure). The slope of the straight line has been the subject of debate ever since Farmer proposed his criterion more than 30 years ago (Reference 2). A slope of -1 reflects a " risk-neutral" attitude, which means that, if the consequencesincrease by an order of magnitude, the corresponding frequency decreases by an order of magnitude also. The Dutch nuclear regulatory body has adopted Farmer curves for prompt fatalities with a slope of -2, thus exhibiting a strong risk-aversive attitude. Can the Regulatory Guide 1.174 Process Benefit from the 10 CFR 50.59 Process? Yes. Regulatory Guide 1.174 requires that AU increases in frequency must be reviewed by the staff, that is, Figure 2 does not have a region in which the change in frequency is so small that subjecting it to staff review would be a waste of everyone's resources. The following interesting comment accompanies this figure in Regulatory Guide 1.174:
"The analysis will be subject to increased technical review and management attention as indicated by the darkness of the shading of the figure."
Clearly, this " attention" is significant when either the goal is exceeded or when the change in frequency approachesthe limit of 104per reactor-year. An example of a region in which no review would be required is shown in Figure 4. What Shouldbe Done? The staff should evaluate the feasibility of employing frequency-consequencecurves for the whole spectrum of accidents (Classes 1 through 9). In particular, this evaluation should include: 1. the definition of appropriate metrics for the consequences (I used iodine releases as an example only).
- 2. the definition of appropriate Farmer curves. 3. the definition of appropriate guidelines as to what constitutes a " minimal impact" on the frequency-consequence curves that would be applied for the whole spectrum of accidents and for all plants.
This effort should be coordinated with the current staff activities on possible revisions to the Safety Goal Policy Statement. For example, in its report to the Commission on " Elevation of CDF to a Fundamental Safety Goal and Possible Revision of the Commission's Safety Goal Policy Statement" dated May 11,1998, the ACRS recommended that the staff evaluate the possibility of using three regions for some of the objectives. If the Policy Statement is revised to include three regions for LERF, then two Farmer curves should be defined for releases from Class 1 through 9 accKients, so that three regions would also be defined on the F-C plane. 106
i 1 5 t What are the Benefits? I realize that what I am proposing would be a significant change in the regulations. I do believe, however, that the benefits warrant this bold step. Specifically,
- 1. The processes of 10 CFR 50.59 and Regulatory Guide 1.174 would be consistent.
- 2. Debates about "who owns" the operating margins would be unnecessary.
- 3. A significant first step would be taken toward making Part 50 risk-informed.
- 4. A significant first step would be taken toward establishing the basis for the determination of rational performance criteria.
- 5. The regulatory system would be more rational and much less intrusive than the current one.
References
- 1. U.S. Nuclear Regulatory Commission, NUREG/CR-0603, "A Risk Assessment of a Pressurized Water Reactor for Class 3-8 Accidents," A. Busiik et al., October 1979.
- 2. F. R. Farmer, " Siting Criteria - A New Approach," presented at the Conference on Containment and Siting of Nuclear Power Plants, Intemational Atomic Energy Agency, !
Vienna,1967 (reproduced in Nuclear Safety, vol. 8, pp. 539-548). ! l f 107
6 Frequency of R > R* (yr-1) imm
! Iodine Group ==rr
....,- .... m sme 8
- 8.gespek
, - . ' *% g 7 -
- A - e k m.u A.i
~
+ " **
im* s
^
)
-0 0 ^ -;
m- ; ;; -
. cw a.es. .
4s s !
- s s. N.
1.es-er g 3, \g s.es* , i 0
~
m . . . . . . . . . . . . . . . . . . . . . . . . . 1AS-06 1.eS 46 1.es-es 1.es-es i.es-41 t.eg+es Release Fraction , Figure 1. Example of F-C curves for iodine release (Fig. 2.7 of NUREG-1150, vol.1) { J l 108
7 t u. M ua J
<a 4
10
- g- 5+.-t?Wy REGION 5IIi 2 %,:y.sy 2%;s .i
, ' < 3 ".;
i_ ::.'T? d' nI , 10 ' 2.. ' - s, REGION III d 10 10-5 LERF
- Figure 2. Acceptance Guidelines for Large Early Release Frequency (LERF), as given in RG 1.174.
109
I 8 Logf i Class 1-8 Accidents Severe Accidents Logc Fig. 3. Example of a Farmer Curve. t u. W. ua J 4 10
'///
< :%hkW.
/ . REGION
}
II:o ;AO'\Qy m.t., 10 4 -
' , , ' ,/ -
f-
/ ,RNGION IV;[ / REGION i
, III 4 LERF
- 10 10-5 Figure 4. Acceptsace guidelines for LERF including a region (IV)in which no prior review is required.
i 110 1 I
e ucg'g UNITED STATES 8 n NUCLEAR REGULATORY COMMISSION U I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsmNGTON, D. C. 20555 July 20, 1998 The Honorable Shirley Ann Jackson Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Chairman Jackson:
SUBJECT:
DRAFT SUPPLEMENT 1 TO NUREG 1552, " FIRE BARRIER PENETRATION SEALS IN NUCLEAR POWER PLANTS" During the 454th meeting of the Advisory Committee on Reactor Safeguards, July 8-10,1996^, we met with representatives of the NRC staff, the Nuclear Energy Institute (NEI), and the Nuclear Information and Resource Service (NIRS) to discuss Supplement 1 to NUREG-1552, We also had the benefit of the documents referenced. RECOMMENDATIONS
- 1. Supplement 1 to NUREG-1552 should be issued for public comment.
- 2. We agree with the Commission's direction in the June 30,1998 Staff Requirements Memorandum that Section Ill. M of Appendix R to 10 CFR Part 50 should be amended to eliminate the requirement that penetration seal designs should utilize only noncombustible materials.
- 3. Steps should be taken to refine NRC inspections of fire barrier penetration seals and to allow licensees to focus their penetration seal programs based on risk.
DISCUSSION Defense in depth against the effects of fire in nuclear power plants includes division of the plants into ' fire areas' separated by fire barriers. The fire barriers are intended to prevent the spread of fire from one area to adjacent areas. The fire barriers are, of course, penetrated by piping, cable trays, and the like. Spread of fire along these penetrations is prevented by fire banier penetration seals. The average number of fire barrier penetration seals per nuclear plant unit is about 3000, and a single unit can have up to 10,000 seats. Fire barrier penetration seals have been receiving much public attention in recent months. Indeed, many years ago, shortly after NRC upgraded its regulations and requirements for fire protection at nuclear power plants in response to the 1975 event at the Browns Ferry Nuclear 111
2 Power Plant, licensees were having substantial difficulties with the installation and maintenance of penetration seals. Over the last several years, the NRC staff has conducted three major reviews ofinstalled penetration seals, as well as inspections of the qualification of penetration seals for service in nuclear power plants. These major, generic examinations were in addition to the ongoing monitoring of seals by the NRC inspectors at individual plants. At the same time, licensees have greatly upgraded their penetration seal programs to ensure that they comply with the regulatory requirements. Findings of the staff investigations and reviews are summarized in NUREG-1552. More recently, the NRC staff has reviewed licensee event reports and inspection reports on deficiencies in fire barrier penetration seals. The staff has found that the frequency of problems with the seals are low and appear to be decreasing with time. These findings suggest that, on average, no more than one seal is deficient each year at a nuclear plant at present. It is unlikely that any justifiable, additional efforts by licensees or the NRC staff could improve the availability of the penetration seals. Results of this recent review are included in the Draft Supplement 1 to NUREG-1552. Concems have arisen because the material used commonly for fire barrier penetration seals in both nuclear and nonnuclear applications is " combustible" according to a particular test that has little to do with actual service applications of the seals. Typically, these seals are required to prevent the spread of fire for three hours. Realistic tests of both the material and the configuration of the seal in use demonstrate that fire barrier penetration seals at nuclear power plants are capable of fully meeting the service requirements for fire protection. Data from real fires at U. S. nuclear power plants do not suggest any additional concems. Therefore, there is no need for the existing regulatory requirement for using only noncombustible materials for fire penetration seals. It is sufficient to require demonstration that the material in its particular configuration meets the functional requirements for fire protection. It is clear that the NRC staff and the licensees, overall, have the issues of fire barrier penetration seals well in hand. Though specific events at specific plants still can be expected to occur, there are no widespread or potentially generic problems of safety significance associated with the seals or the seal materials now in use. The efforts of the staff and the licensees on fire penetration seals have been successfulin addressing problems of the past. At present, these efforts are out of proportion to the risk significance of fire barrier penetration seals. These efforts do amount to significant burdens on the !icensees and significant drains on NRC resources. Fire risk assessment can be used as the basis for focusing these efforts on risk significant fire barrier penetration seals with no reduction in the protection of public health and safety. That is, inspections of the fire barrier penetration seals in a nuclear power plant could be based on risk in a manner similar to the risk-based inspection of piping systems. 112
3 To refine the programs now in place to deal with fire barrier penetration seals and other fire protection issues based on risk will require the availability of reliable fire risk assessment tools that treat the nuclear power plant in sufficient detail. Data on the reliability and performance of seals even when degraded will also be needed This effort may well require that the NRC undertake research to further develop its analytical tools and databases. Sincerely, R. L. Seale Chairman q References.
- 1. U. S. Nuclear Regulatory Commission, NUREG-1552, Supp.1," Fire Barrier Penetration Seals in Nuclear Power Plants," Draft Report for Comment, June 1998.
- 2. Memorandum dated June 30,1998, from John C. Hoyle, Secretary of the Commission, j to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
Staff Requirements: SECY 98-058 - Development of a Risk-Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants. 113
L-
/ o g UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION
{ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wasuincrow,o.c. oses 1. July 21,1998 Mr. L Joseph Callan Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Mr. Callan:
SUBJECT:
PROPOSED FINAL SAFETY EVALUATION OF THE BWR VESSEL AND INTERNALS PROJECT, BWR PRESSURE VESSEL SHELL WELD INSPECTION RECOMMENDATIONS (EWRVIP-05) REPORT During the 454th meeting of the Advisory Committee on Reactor Safeguards, July 8-10,1998, we reviewed the proposed final safety evaluation of the BWR Vessel and intemals Project (BWRVIP-
- 05) report concerning industry recommendationsfor reducing the scope of inservice inspection of BWR reactor vessel welds. During our review, we had the benefit of discussions with representatives of the NRC staff and of the documents referenced.
- g. n Conclusions
. We endorse the staff's recommendation that licensees be granted permanent relief from inservice inspection requirements for volumetric examination of BWR circumferential reactor pressure vessel welds if the licensee can demonstrate that the generic evaluation performed by the staff is applicable to its vessel.
. We concur with the staff's request that the BWRVIP provide a plan for followup analyses to determine more realistic estimates of the frequency of axial weld failures caused by cold- ,
overpressure events and propose appropriate technical approaches to address this issue. j Discussion in our September 10,1997 letter, we recommended that the staff review the BWRVIP-05 report , using the risk-informed process in Regulatory Guide 1.174, "An Approach for Using Probabilistic l Risk Assessmentin Risk-informedDecisions on Plant-Specific Changes to the Current Licensing Basis." We also recommended that additionalefforts be taken to address uncertainties associated with the industry and staff analyses,in particularthose associated with flaw size distributions and I the sequences that could lead to vessel challenges. In addition, we recommended that the staff l considerthe value of partialinservice inspechon of welds. In response, the staff and the industry estimated the frequency of vessel challenges through studies of potential precursor events, used recent research results from examination of actual pressure vessel welds to update flaw size I
- distributions, and performed additional probabilistic fracture mechanics analyses.
115
F I I The staff and industry studies confirmed that the failure frequency of axial welds during cold over-pressure events is much greater than the failure frequency for circumferential welds. The circumferential weld failure frequency is below the criteria specified in Regulatory Guide 1.154, j
" Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," for pressure vessel integrity. This finding supports the conclusion that inservice inspection of circumferential welds is not necessary during the current license term for operating BWRs.
The computed failure frequency of the axial welds does not meet the criteria of Regulatory Guide 1.154. This assessmentis based on end-of-life neutron fluence levels that will not occur for many years and includes a number of additional conservatisms, and hence, failure of the axial welds is not a near-term safety concem. The staff requested additional analyses from the BWRVIP to obtain more realistic estimates of axial weld failure frequency. The studies performed by the staff and the industry also demonstrate that inservice inspection of the axial welds is ineffective in reducing the likelihood of vessel failure due to fabrication flaws. An inspection program for these welds consistent with the intent of ASME Section XI, however, does provide assurance that service-induced degradation mechanisms will be detected and is an important element of defense in depth. Dr. William J. Shack did not participate in the Committee's deliberation regarding this matter. Sincerely, R. L. Seale Chairman References-
- 1. Letter dated September 10,1997, from R. L. Seale, Chairman, ACRS, to L. Joseph Callan. 1 Executive Director for Operations, NRC,
Subject:
Boiling Water Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05).
- 2. Memorandum dated June 3,1998, from Frank J. Miraglia, Office of Nuclear Reactor Regulation, NRC, to Robert L. Seale, Chairman, ACRS,
Subject:
Transmittalof NRC Staff's Draft Safety Evaluation of the "BWR Vessel and Intemals Project, BWR Reactor Pressure ) Vessel Shell Weld Inspection Recommendations (BWRVIP-05)" Report. I 116
1 km K trug'o j
,, UNITED STATES I
/ ,, NUCLEAR REGULATORY COMMISSION , y, ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o WASHINGTON, D. C. 20555 July 22,1998 MEMORANDUM TO: L. Joseph Callan Executive Director for n FROM: John T. Larkins, E ecut' irector i Advisory Committee on Reactor Safeguards
SUBJECT:
DRAFT REGULATORY GUIDE DG-4005, " PREPARATION OF SUPPLEMENTAL ENVIRONMENTAL REPORTS FOR APPLICATIONS TO RENEW NUCLEAR POWER PLANT OPERATING LICENSES" During the 454th meeting of the Advisory Committee on Reactor Safeguards, July 8-10, 1998, the Committee decided not to review the subject document, and has no objection to issuing it for public comment. Reference U. S. Nuclear Regulatory Commission, Draft Regulatory Guide DG-4005," Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses," June 1998. cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR J. Roe, NRR M. Federline, RES D. Cleary, RES 117
/ %o UNITED STATES 8 NUCLEAR REGULATORY COMMISSION y, I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHWGToN, D. C. 20656 July 23,1998 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
REPORT ON THE SAFETY ASPECTS OF THE WESTINGHOUSE ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE AP600 PASSIVE PLANT DESIGN During the 454th meeting of the Advisory Committee on Reactor Safeguards, July 8-10,1998, we completed our safety review nf the Westinghouse Electric Company application for certification of its AP600 passive plant design. This report is intended to fulfill the requirement of 10 CFR 52.53 that "the ACRS shall report on those portions of the application which concem safety." During our review, we had the benefit of discussions with representatives of Westinghouse and its consultants, and the NRC staff. We also had the benefit of the documents referenced. AP600 Apphcation On June 26,1992, Westinghouse tendered its application to the NRC for certification of the AP600 design. This application was submitted in accordance with Subpart B, " Standard Design Certifications," of 10 CFR Part 52, "Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants," and Appendix O, " Standardization of Design: Staff Review of Standard Designs." The application was docketed on December 31,1992, and Essigned Docket Number 52-003. The application consists of the AP600 Standard Safety Analysis Report (SSAR), the Tier 1 Material, and the probabilistic risk assessment (PRA). On June 26,1992, Westinghouse submitted the SSAR and the PRA. In December 1992, Westinghouse submitted the Tier 1
. Material, which contains inspections, tests, analyses, and acceptance criteria (ITAAC) and Tier 1 design descriptions. Design certification is sought for the power generation complex, cxcluding those elements and features considered site-specific. All safety-related structures, systems, and components (SSCs) are located on the nuclear island and are to be included in the design certification.
119
F i 2 Three aspects of the plant design (i.e., instrumentation and control (l&C) systems, human factors engineering, and some piping) will be completed by the combined license (COL) i applicant using the design processes described in the SSAR and ITAAC. The staff issued a Draft Safety Evaluation Report (DSER) on November 30,1994, a supplement to the DSER in April 1996, and an Advance Final Safety Evaluation Report on May l 2,1998. Our activities related to the review of the AP600 design are listed in the Attachment. As a result of our review, we issued three interim letters identifying several issues. The I resolution proposed by Westinghouse to these issues is acceptable, pending staff review and approval. AP600 Design Desenption The AP600 plant is designed for use at either single-unit or multiple-unit sites. The scope of the J design is complete except for site-specific elements. The AP600 design has a nuclear steam supply system rating of 1933 MWt, with an electrical output of at least 600 MWe, The plant has a design objective of 60 years without a planned replacement of the reactor vessel. The design does provide, however, for the replacement of other major components, including the steam generators. The primary objective of the AP600 design is to meet safety requirements and goals defined for advanced light-water reactors with passive safety features as specified in the Electric Power Research Institute Utility Requirements Document. An additional objective is to provide a greatly simplified plant with respect to design, licensing, construction, operation, inspection, and maintenance. . I The plant arrangement consists of five principal structures; the nuclear island, the turbine building, the annex building, the diesel generator building, and the radwaste building. The nuclear island, which includes all safety-related or seismic Category I structures, is ! designed to withstand the effects of natural phenomena and postulated events. It consists of a containment building, a concrete shield building, and an auxiliary building, which are described below.
. The containment building consists of a free-standing steel containment vessel which has a design pressure of 45 psig and associated intemal structures. The vessel performs the function of containing the release of radioactivity to the atmosphere foll> wing postulated design-basis accidents. The vesselis also part of the passive containment cooling system. . ' The shield building comprises the structure and annulus area that surrounds the 1 containment building. In the event of an accident, the passive containment cooling system releases water that runs down the outside of the containment vessel to enhance heat removal. . The auxiliary building is designed to provide protection and separation for the seismic Category 1 mechanical and electrical equipment located outside the containment 120
r 3 building. The building also provides protection for safety-related equipment against the consequences of intemal or extemal events. The main control room, Class 1E l&C systems, Class 1E electrical systems, and reactor fuel handling area are contained in - the auxiliary building. The turbine building houses the main turbine generator and associated fluid and electrical systems. The annex building includes the health physics area, the technical support center, access control, and personnel facilities. The diesel generator building houses two diesel generators and their associated support systems. The radwaste building contains facilities for the handling, processing, and storing of radioactive wastes. The overall plant arrangement utilizes building configurations and structural designs to minimize the building volumes and quantities of bulk materials (concrete, structural steel, rebar) , consistent with safety, operational, maintenance, and structural needs. The plant arrangement provides separation between safety and nonsafety equipment and systems to preclude adverse interactions among them. Separation between redundant safety equipment and systems provides confidence that the safety functions can be performed. In general, this separation is provided by concrete walls. The ITAAC program is intended to ensure that the plant, when built, conforms to the design parameters and assumptions that existed at the time of design certification. For example, the efficacy of the passive emergency core cooling system depends on the flow resistances of piping segments, relief valves, and other components. The flow resistances will be measured in the as-built plant to ensure that they conform with the values derived and validated by the test and analysis program. Safety Enhancement Features The AP600 design contains many features that are not found in current operating plants. For example, a variety of engineering and operational improvements provide additional safety margins and comply with the Commission's Severe Accident, Safety Goal, and Standardization Policy Statements. Unique features of the AP600 design include an improved reactor core design, a large reactor vessel, a large pressurizer, an in-containment refueling water storage tank (IRWST), an automatic depressurization system, a digital microprocessor-based l&C system, hermetically sealed canned rotor coolant pumps mounted to the steam generator, and increased battery capacity. The AP600 design represents a significant departure from previous commercial nuclear reactor technology in that it places more dependence on passive systeros for accident response. Passive systems depend on gravity, condensation, and small pressure differences to prevent or mitigate damage to the core and to ensure containment of radioactive fission products in the ovent of accidents. Active systems, on the other hand, employ flow loops and pumps that require electrical or other sources of motive power. The performance of active systems is, in general, better known because of existing test data and extensive operating experience. Passive systems, although not tested under full-scale conditions, are more likely to ensure safety functions, especially under conditions where extemal or emergency motive power could be compromised. 121
r I 4 l [ l The AP600 l&C systems are significantly different from those in current operating plants. The i primary differences result from using software-based digital systems with multiplexed and fiber optics data links in place of the analog systems. The use of digital systems with multiplex and fiber optics data links reduces the amount of cabling in the plant, thereby reducing configuration complexity and fire hazards. The AP600 design does not require Class 1E electrical power except that provided by the Class 1E dc batteries and their inverters. This feature significantly reduces the complexity of the plant electrical systems and the reliance on safety-grade diesel generators. The AP600 plant includes an innovative security plan which features the use of defensive capabilities at various vital area access points. This feature results in elimination of the protective area boundary a'nd associated security attributes used at current operating nuclear power plants. AP600 Test and Analysis Proaram Westinghouse conducted an extensive test and analysis program, utilizing separate-effects and integral-system facilities both to investigate the behavior of the AP600 passive safety systems and to develop a database for validation of the computer codes used to perform accident and transient analyses. Key aspects of the test and analysis program include:
. Core Makeup Tank (CMT) Test Program to characterize the CMT over an extended range of thermal-hydraulic conditions.
. Automatic Depressurization System (ADS) Test Program, both to characterize the steam flow through the IRWST sparger and to test the thermal-hydraulic behavior of the ADS piping network.
- Passive Residual Heat Removal (PRHR) System Test Program to generate data for design and characterization of the AP600 PRHR heat exchanger.
. Oregon State University Advanced Plant Experiment (APEX) Test Program to obtain integral-systems data for code validation; emphasis was placed on low-pressure and long-term core cooling behavior for design-basis, small-break loss-of-coolant accidents (LOCAs).
. SPES-2 High-Pressure, Full-Height Integral-Systems Test Program to obtain integral-systems data for code validation; the particular focus was on accident progression from initiation to establishment of stable IRWST injection.
. Passive Containment Cooling System Test Program to obtain integral-systems test data on the thermal-hydraulic performance of this system to support code validation.
This extensive test and analysis program was necessary to validate the accident analysis codes applied to new, passive emergency core cooling systems for which there is not a significant experience base. The accKient analysis codes used by Westinghouse included: 122
y 5 l . LOFTRAN/LOFTTR2 for analyses of non-LOCA transients l . NOTRUMP for evaluation-model analyses of small-break LOCAs l . WCOBRA/ TRAC for best-estimate analyses of large-break LOCAs !
- WCOBRA/ TRAC for analyses of long-term core cooling e WGOTHIC for design-basis accident analyses of the containment l
To ensure that the test and analysis program adequately addressed important phenomena with r spect to the passive systems and that the results would scale to the prototype size, Westinghouse developed a phenomena identification and ranking table and performed a scaling analysis for both the primary coolant system and the containment. In addition, the NRC staff performed confirmatory experimental and analytical programs in support of the AP600 design certification review. These programs included the integral-systems testing performed at the Japan Atomic Energy Research Institute ROSA-AP600 facility, and follow-on testing performed at the Oregon State University APEX facility. The NRC stiff also performed confirmatory analyses utilizing the NRC codes RELAP-5 and CONTAIN. The results of the staff's programs significantly aided our review of the Westinghouse test and Enalysis program. During the extensive reviews of the Westinghouse test and analysis program, we raised numerous issues. These issues have been documented in our interim letters and meeting minutes. Based on discussions with representatives of Westinghouse and the NRC staff, all of our issues pertaining to the Westinghouse test and analysis program have been adequately r solved. There are, however, a number of issues that arose during our review that, while not directly Effecting the acceptability of the AP600 test and analysis program, should be considered in the context of future design certification reviews. We plan to address these issues in a future letter pertaining to lessons leamed from the AP600 design certification review. Probabilistic Risk Assessment The AP600 design certification application included a PRA, in accordance with regulatory
. rsquirements. This PRA was done well and rigorous methods were used to quantify risk metrics, including core damage frequency (CDF) and large, eariy release frequency (LERF).
Point estimates of the risk metncs are: i 4 CDF = 2 X 10 per reactor year LERF = 2 X 10 4per reactor year These risk metrics are low compared to those estimated for existing nuclear power plants. The l PRA was an integral part of the design process. This contributed significantly to design l modifications, which resulted in the low CDF and LERF. 1 The PRA addressed passive safety systems and software-based digital l&C systems. Qualitative analyses and extensive sensitivity studies were used to compensate for incomplete modeling of these important features of the plant. In addition, the concept of the " focused" PRA 123
I i 6 I was introduced to reduce uncertainties in the estimated performance of passive systems. The objective of the " focused" PRA was to determine whether the goals for CDF and LERF could be met without the support of the nonsafety-ralated systems. The regulatory treatment of ; I nonsafety systems (RTNSS) process was used to impose special requirements on some nonsafety systems to ensure, with high confidence, that they would be available when needed. j
~
For example, Westinghouse used the RTNSS process to impose administrative controls on the availability of the engineered safety feature actuation function of the diverse actuation system in , order to reduce uncertainties associated with the digital system software. The RTNSS process is an excellent example of a good risk-informed and performance-based approach. We applaud the use of the " focused" PRA and the RTNSS process in developing defense-in-depth measures. But, we caution against establishing the practice of comparing the results of
" focused" PRAs with Safety Goals. These Goals apply to a plant as it is designed and operated. Comparison of these Goals with results of analyses, restricted to include only safety f systems, would amount to the imposition of a new goal that does not appear in the Commission's Safety Goal Policy Statement.
Additional Observations Westinghouse's approach for quantifying digital systems software in the PRA is consistent with the guidance in Regulatory Guide 1.152, " Criteria for Digital Computers in Safety Systems in Nuclear Power Plants." This approach provides a method for identifying and assessing design strengths and weaknesses. The AP600 plant will use passive autocatalytic recombiners to maintain hydrogen concentrations below the flammability limit within the containment following design-basis accidents. We agree, in principle, that these devices are improvements over hydrogen recombiners used in existing plants. The COL applicant is responsible for qualifying passive autocatalytic recombiners. The present regulatory requirements for qualifying mechanical { equipment are insufficient to ensure continued passive autocatalytic recombiner operation for the expected duty cycle. The AP600 reactor containment is a steel shell. It has been designed to meet Service Level C of the ASME Boiler and Pressure Vessel Code. The containment meets all regulatory requirements. Testing has shown that steel shell containments are susceptible to catastrophic failure when overpressurized. For the AP600 design, however, under the peak pressure calculated in the Level 2 PRA for severe accident conditions, the probability of failure of the ' containment is estimated to be approximately 0.01. Deformation of the pressurized containment vessel and its interaction with the shield building could also induce leakage and further reduce the likelihood of failure, in any event, we have not been able to identify significant risks associated with possible catastrophic failure modes of the AP600 containment. I Westinghouse has concluded that extemal reactor vessel cooling will prevent core debris from l penetrating the reactor vessel. This conclusion is based on a scenario for degradation of the i core that avoids consideration of direct contact by metallic core debris with the reactor vessel. The NRC staff has concluded that reactor vessel failure is not precluded and has required that l Westinghouse consider ex-vessel core debris interactions. Westinghouse performed these 124
i 7 l evaluations and found that the AP600 containment performs satisfactorily under these severe conditions. ACRS Conclusion Concernina AP600 Desian
- Based on our review of those pomons of the AP600 application which concem safety, we l believe that acceptable bases and requirements have been established to ensure that the AP600 design can be used to engineer and construct plants that with reasonable assurance can be operated without undue risk to the health and safety of the public.
Dr. Thomas S. Kress did not participate in the Committee's deliberation regarding extemal r actorvesselcooling. Dr. Dana A. Powers did not participate in the Committee's deliberation regarding the AP600 source term or the results of Sandia National Laboratories tests on containment structural integrity and on environmental qualification of passive autocatalytic recombiners. Dr. George Apostolakis did not participate in the Committee's deliberation regarding the AP600 passive system reliability assessment or the analyses performed by the Idaho Engineering and Environmental Laboratory conceming the use of the WCOBRA/ TRAC code and extemal reactor vsssel cooling. Sincerely, R.L.Seale Chairman References
- 1. U.S. Nuclear Regulatory Commission, " Advance Final Safety Evaluation Report Related to the Certification of the AP600 Design," May 1998 (Predecisional information).
- 2. U.S. Department of Energy Report DE-AC03-90SF18495, dated June 26,1992, prepared by Westinghouse Electric Corporation, "AP600 Standard Safety An'alysis Report," updated through Revision 23 (issued May 18,1998).
- 3. U.S. Department of Energy Report DE-AC03-90SF18495, dated June 26,1992, j prepared by Westinghouse Electric Company, "AP600 Probabilistic Risk Assessment," !
updated through Revision 11 (issued March 1998).
- 4. U.S. Department of Energy Report DE-AC0190SF18495, December 1992, "AP600 Tier 1 Material," updated through Revision 5 (issued May 7,1998). !
- 5. Letter dated February 19,1998, from R. L. Seale, Chairman, ACRS, to L. Joseph i Callan, Executive Director for Operations, NRC,
Subject:
Interim Letter on the Safety 125
8 i Aspects of the Westinghouse Electric Company Application for Certification of the AP300 Plant Design.
- 6. Letter dated April 9,1998, from R. L. Seale, Chairman, ACRS, to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
The Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design-Interim Letter 2.
- 7. Letter dated June 15,1998, from R. L. Seale, Chairman, ACRS, to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
The Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design-Interim Letter 3.
- 8. U.S. Nuclear Regulatory Commission, Draft NUREG-1512, " Draft Safety Evaluation Report Related to the Certification of the AP600 Design," November 1994
- 9. U.S. Nuclear Regulatory Commission, Supplement to the " Draft Safety Evaluation i Report Related to the Certification of the AP600 Design," April 1996.
- 10. Westinghouse Electric Corporation, WCAP-14812, Revision 2, " Accident Specification and Phenomena Evaluation for AP600 Passive Containment Cooling System," April 1998 (Proprietary).
- 11. Westinghouse Electric Corporation, WCAP-14845 Revision 3," Scaling Analysis for AP600 Containment Pressure During Design Basis Accidents," March 1998 (Propnetary).
- 12. Westinghouse Electric Corporation, WCAP-14326, Revision 3, " Experimental Basis for the AP600 Containment Vessel Heat and Mass Transfer Correlations," May 1998 (Propnetary).
- 13. Westinghouse Electric Corporation, WCAP-14135, Revision 1, " Final Data Report for PCS Large-Scale Tests, Phase 2 and Phase 3," April 1997 (Proprietary).
- 14. Westinghouse Electric Corporation, WCAP-14382, Revision 0, "WGOTHIC Code Description and Validation," May 1995 (Proprietary).
- 15. Westinghouse Electric Corporation, WCAP-14407, Revision 3, "WGOTHIC Application to AP600," April 1998 (Proprietary).
-16. Office of Nuclear Regulatory Research (RES) Report, RPSB-98-04, "Phenomenology Observed in the AP600 Integral Systems Test Programs Conducted in the ROSA-AP600, APEX, and the SPES-2 Facilities," D. Bessette, RES, M. DiMarzo, University of Maryland, P. Griffith, Massachusetts Institute of Technology, April 1998.
- 17. Letter dated July 1,1998, from B. McIntyre, Westinghouse, to NRC: Attention: J. Larkins, ACRS, transmitting Response to ACRS Request for NOTRUMP Break Area
. Sensitivity Study.
]
- 18. Letter dated July 2,1998, from B. McIntyre, Westinghouse, to NRC: Attention: J.
La: tins, ACRS,
Subject:
Responses to ACRS Reactor Coolant System issues.
- 19. Lette dated June 25,1998, from B. McIntyre, Westinghouse, to NRC: Attention: J.
Larkins, ACRS,
Subject:
Closure of ACRS Thermal Hydraulic Subcommittee items for June 11-12,1998 Meeting.
Attachment:
Chronology of the ACRS Review of the Westinghouse Application for the AP600 Passive Plant Design Certification ; i i i 126 ,
l 1 ATTACHMENT CHRONOLOGY OF THE ACRS REVIEW OF THE WESTINGHOUSE APPLICATION FOR THE AP600 PASSIVE PLANT DESIGN CERTIFICATION The extensive ACRS review of the AP600 design and its interactions with representatives of the NRC staff and Westinghouse are discussed in the minutes of the following ACRS meetings. The questions raised by ACRS members during meetings which were not formally documented in ACRS reports and letters were answered during subsequent discussions. ACRS MEETING / DATES SUBJECT Thermal Hydraulic Phenomena Proposed Commission Paper on Need for Full-12/17/91 Height, Full-Pressure Integral System Testing of AP600 Design Thermal Hydraulic Phenomena Integral System Testing Requirements for AP600 3/3/92 Design Thermal Hydraulic Phenomena integral System Testing Requirements for AP600 6/23-24/92 Design Thermal Hydraulic Phenomena Office of Nuclear Regulatory Research (RES) 3/4-5/93 RELAP5/ MOD 3 Code Thermal Hydraulic Phenomena Westinghouse Test and Analysis Program (TAP) 7/22-23/93 Thermal Hydraulic Phenomena TAP - Oregon State University APEX Test Facility 9/21/93 Thermal Hydraulic Phenomena RES - ROSA-V (ROSA-AP600) Confirmatory Test 10/28/93 Program Thermal Hydraulic Phenomena RES - RELAP5/ MOD 3 Code 1/4-5/94 Thermal Hydraulic Phenomena TAP - Core Makeup Tank Test Facility, 3/15-16/94 Passive Containment Cooling System Thermal Hydraulic Phenomena TAP - WCOBRA/ TRAC Code 5/18-19/94 127
i l l 2 Thermal Hydraulic Phenomena RES - Confirmatory Test Programs 8/25-26/94 W Standard Plants Designs Overview and General Description of the AP600 Plant 1/11/95 Design Thermal Hydraulic Phenomena TAP - WCOBRA/ TRAC Code 2/15-16/95 Thermal Hydraulic Phenomena RES - Phenomena identification and Ranking 3/27-28/95 Table (PIRT) for RELAP5 Code Thermal Hydraulic Phenomena TAP - Passive Containment Cooling System 3/29-30/95 W Standard Plant Designs Commission Paper on Status of Ten Key 5/31/95 Technical and Policy issues Thermal Hydraulic Phenomena Qualification Document for the WCOBRA/ TRAC Code 7/26-27/95 Thermal Hydraulic Phenomena Qualification Document for the WCOBRA/ TRAC Code 1/18-19/96 Thermal Hydraulic Phenomena RES Program for Demonstrating Adequacy of the 2/22-23/96 RELAP5/ MOD 3 Code to Assess Behavior of AP600 Design Thermal Hydraulic Phenomena TAP - Overview 5/9-10/96 Severe Accidents Probabilistic Risk Assessment of Severe 6/5/96 Accidents W Standard Plant Designs SECY-96-128, " Policy and Key Technical Issues 7/19/96 Pertaining to the AP600 Design" 433rd ACRS Meeting SECY-96-128, " Policy and Key Technical issues 8/8/96 Pertaining to the AP600 Design" ACRS Report issued 8/15/96 W Standard Plant Designs Chap. 4: Reactor 12/4/96 Chap. 5: Reactor Coolant System and Connected Systems Chap. 9: Auxiliary Systems Chap.11: Radioactive Waste Management 128
p t 3 l Thermal Hydraulic Phenomena TAP - Scaling and PIRT Closure Report 12/18-19/96 Thermal Hydraulic Phenomena RES Program for Demonstrating Adequacy of the 2/12-14/97 RELAP5/ MOD 3 Code to Assess Behavior of AP600
)
i Design Thermal Hydraulic Phenomena RES - ROSA-AP600 Confirmatory Test Program 2/19/97 Thermal Hydraulic Phenomena TAP - Long-Term Cooling with WCOBRA/ TRAC 3/28/97 Code i i 442nd ACRS Meeting AP600 Containment Spray System 6/13/97 ACRS Report issued 6/17/97 Thermal Hydraulic Phenomena TAP - NOTRUMP Small-Break LOCA Code 7/29-30/97 Thermal Hydraulic Phenomena TAP - Passive Containment Cooling System 9/29-30/97 Thermal Hydraulic Phenomena TAP - PIRT; Scaling of Reactor Coolant System; 12/9-10/97 NOTRUMP Code Thermal Hydraulic Phenomena TAP - WGOTHIC Containment System Code 12/11-12/97 Advanced Reactor Designs Chap. 7: Instrumentation and Controls 2/3-4/98 Chap. 8: Electrical Power Chap.13: Conduct of Operations , Chap.18: Human Factors Engineering i 448th ACRS Meeting TAP 2/5/98 Chap. 1: Introduction and General Discussion Chap. 4: Reactor Chap. 5: Reactor Coolant System and Connected Systems Chap. 7: Instrumentation and Controls Chap. 8: Electncal Power Chap.11: Radioactive Waste Management Chap.13: Conduct of Operations Chap.18: Human Factors Engineering Interim fCRS letter issued 2/19/98 Advanced Reactor Designs Chap. 2: Site Characteristics 3/30 - 4/1/98 Chap. 9: Auxiliary Systems l 129
l l \ 4 Chap.10: Steam and Power Conversion Chap.12: Radiation Protection Chap.13: Conduct of Operations (Security) Chap.15: Accident Analyses 451st ACRS Meeting TAP 4/2/98 Chap. 2: Site Characteristics Chap. 9: Auxiliary Systems Chap.10: Steam and Power Conversion Chap.12: Radiation Protection Chap.13: Conduct of Operations (Security) Chap.15: Accident Analyses Interim ACRS Letter 2 issued April 9,1998 Thermal Hydraulic Phenomena TAP - Primary Coolant System 5/11-12/98 Advanced Reactor Designs Chap. 1: Introduction and General Discussion 5/13-15/98 Chap. 6: Engineered Safety Features Chap.14: Initial Test Program Chap.16: Technical Specifications Chap.17: Quality Assurance Levels 2 and 3 PRA Regulatory Treatment of Nonsafety Systems 453rd ACRS Meeting TAP 4 6/3/98 Chap. 3: Design of Structures, Components, Equipment, and Systems Chap. 6: Engineered Safety Features Chap. 9: Appendix A - Fire Protection Analysis Chap.14: Initial Test Program Chap.16: Technical Specifications Chap.17.: Quality Assurance l PRA Interim ACRS Letter 3 issued June 15,1998 Thermal Hydraulic Phenomena TAP - Passive Containment Cooling System 6/11-12/98 Advanced Reactor Designs ITAAC; Level 1 PRA; Adverse Interaction 6/17-18/98 Evaluation Report; and Containment Spray System Advanced Reactor Designs TAP and Responses to ACRS Questions 7/7/98 l l 130
1s:o i
#o g UNITED STATES i 8 o NUCLEAR REGULATORY COMMISSION $ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g WASHINGTON, D. C. 20555 es, , s *'
i I July 22,1998 1 I MEMORANDUM TO: L Joseph Callan l Executive Director f o 1 I FROM: John T. Larkins, xect e Director I Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED REVISION TO 10 CFR 50.65(a)(3) OF THE { MAINTENANCE RULE TO REQUIRE LICENSEES TO PERFORM j SAFETY ASSESSMENTS l During the 454th meeting of the Advisory Committee on Reactor Safeguards, July 8-10, 1998, the Committee decided to review the proposed final revision to 10 CFR 50.65(a)(3) following reconciliation of public comments. The Committee has no objection to issuing the l subject document for public comment. '
Reference:
SECY-98-165, dated July 2,1998, for the Commissioners, from L. Joseph Callan, Executive ! Director for Operations,
Subject:
Proposed Revision to 10 CFR 50.65(a)(3) of the Maintenance Rule to Require Licensees to Perform Safety Assessments. cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO M. Federfine, RES S. Collins, NRR S. Black, NRR l 131
I ma teg g%g UNITED STATES o NUCLEAR REGULATORY COMMISSION { .I ADVISORY COMMITTEE ON REACTOR MFEGUARDS WASHINGTON, D. C. 20555 July 24,1998 i i Mr. L. Joseph Callan - Executrve Director for Operations U.S. Nuclear Regulatory Commission ! Washington, DC 20555-0001 l
Dear Mr. Callan:
SUBJECT:
GENERAL ELECTRIC NUCLEAR ENERGY EXTENDED POWER UPRATE PROGRAM AND MONTICELLO NUCLEAR GENERATING PLANT POWER LEVEL INCREASE REQUEST During the 453rd and 454th meetings of the Advisory Committee on Reactor Safeguards, June 3-5 and July 8-10,1998, we reviewed the General Electric Nuclear Energy (GENE) program associated with extended power uprates for operating boiling water reactors (BWRs), and the application by > the Northem States Power Company (NSP) for a power level increase for the Monticello Nuclear Generating Plant. Our Subcommittee on Thermal-Hydraulic Phenomans held a meeting on June 2,1998, to review this matter. During our review, we had the benefit of discussions with representatives of GENE, NSP, and the NRC staff. We also had the benefit of the documents referenced. In 1991, GENE initiated a power uprate program to support BWR plant licensees for increasing rated core power by up to 5 percent. In 1992, we reviewed the initial GENE power uprate program and the application by the Detroit Edison Company for a power level increase for the Fermi nuclear power plant, Unit 2. In our September 17,1992 report, we endorsed the GENE generic program associated with the 5 percent power level uprates and concluded that a 5 percent uprate did not pose a significant increase in risk. It was recognized that any power uprate will in some way erode safety margins and that, although 5 percent uprates were acceptable for all BWRs, any uprates beyond that should be given additional review and justification. in 1995, GENE initiated the " extended" power uprate program. The word " extended"is used to dishnguish this program from the initial power uprate program. The extended uprate program will address additional power uprates greater than 5 percent and up to 20 percent of rated core power. Licensees are to make individual decisions on the magnitude of power uprates. The Monticello Nuclear Generating Plant is the lead plant for the extended power uprate program. l NSP submitted an application for a power level increase of 6.3 percent for the Monticello Plant. l This would increase the current core power level of 1670 MWt to 1775 MWt. In its safety 133
2 evaluation, NSP performed accxient analyses using a core power level of 1880 MWt, which is 12.6 percent above the current core power limit and is double the requested core power level increase. The 6.3 percent power level increase requested and the 12.6 percent power level increase analyzed are achieved with an increase in the steam flow rate but without an increase in maximum allowable core flow or the current maximum operating vessel dome pressure and temperature. The core radial power profile is flattened and the high-pressure turt>ine is replaced with one that accommodates the increased steam flow. l The application by NSP for the Monticello 6.3 percent power uprate utilized the general guidance developed by GENE (ELTR1) and also referenced the GENE generic evaluations (ELTR2). Therefore, any decision on granting the requested uprate has to be accompanied by a decision on the acceptability of the GENE extended power uprate program and the associated reports. The extended power uprate program generally has the objectives of ensuring that all the current ! regulatory requirements will still be met after the uprate and that sufficient safety margins will still exist. The ELTR1 report provides guidance to licensees on the scope and content of information to be submitted as part of a plant-specific power uprate request. The ELTR2 report contains i generic bounding analyses and equipment evaluations in support of the uprate program. These j reports essentially provide a template for any licensee to follow when applying for a power level 1 increase and provide the opportunity to reference any of the bounding analyses that are applicable to the specific application. The staff reviewed the extended power uprate program and presented its evaluation results in two " position papers" - one dated February 8,1996 for ELTR1, and one dated May 18,1998 for ELTR2. The staff generally endorsed this program, but with qualifications. We agree with the staffs assessment and its qualifications and believe that, if followed, this program will provide the information necessary as a basis for the staff review of and decision on plant-specific power uprate applications. We particularly endorse the staff's requirement that "each applicant report the effects of the proposed uprate on its core damage frequency and frequency of large magnitude radioactive release." We believe that the appropriate process for making decisions related to power uprate applications is that outlined in Regulatory Guide (RG) 1.174 related to requests for changes to the licensing basis. With the addition of an analysis for core damage frequency (CDF); large, early release frequency (LERF); and the changes associated with the uprate (ACDF and ALERF), the power uprate program will provide the information required to utilize the RG 1.174 process, including that associated with all the deterministic analyses made as part of a safety evaluation report. In its Safety Evaluation Report (SER), the staff concludes that after the 6.3 percent core power uprate, the Monticello plant meets all the regulatory requirements and preserves appropriate margins. Thus, the submittal meets the requirements for adequate protection. Although the extended power uprate program and the Monticello application preceded by several years the issuance of RG 1.174, significant risk information was provided by NSP in support of the review. The probabilistic risk assessment (PRA) information submitted was based on the 134
3 licensee's individual plant examination (IPE), which included only intemal events. Based on the IPE, the current risk status of the Monticello plant is: CDF = 1.4 x 104/yr, and LERF = 4.5 x 104/yr. The results of the analyses of the 12.6 percent core power uprate are: ACDF = + 2.4 x 104/yr, and ALERF = + 8.6 x 104/yr. These " risk metric" values are within the " allowable change" region specified in the RG 1.174 process The above CDF and LERF absolute values do not include contributions from shutdown End low-power events or from extemal events, nor were they accompanied by any uncertainty analysis. The NSP, however, performed a Fire-Induced Vulnerability Evaluation (FIVE) analysis, e seismic margins analysis, and a shutdown risk analysis from which it would be possible to bound the contributions from these missing elements of the PRA. We believe that an estimate of the effects of the missing PRA elements would not place the Monticello plant outside the " allowable change" region. This should be confirmed by the staff. Provided that the staff confirms that the risk status of Monticello remains in the " allowable change" region specified in RG 1.174, we have the following recommendation: Based on our evaluation of the application and the SER, we agree with the staff's recommendation that the NSP application for a 6.3 percent power level increase for the ) Monticello Plant be approved. We believe this change meets the intent of RG 1.174 to preserve acceptable margins and to limit risk increases to acceptable levels. For future power uprate applications, we have the following recommendations: i
- The staff's recommendation for approval of the power level increase for the Monticello plant is based partly on the IPE that " meets the requirements of GL [ Generic Letter] 88.20."
It is not clear to us that this standard for IPEs is also the appropriate standard for a PRA i on which to base power uprate decisions. A justifiable decision is needed from the staff on the quality standard required for PRAs to assist decisionmaking on power uprate requests. Additional guidance for the applicant is also needed.
. in any future power uprate application, the staff should require that bounding estimates be made for the contributions from any missing elements of the PRA, especially for the contributions from shutdown, low power, and extemal events. . Finally, we are concerned about the concept that seemed to be implied in the application and the staff's review documents that, because better calculations are now possible, greater margins exist. The margir; is inherent in the design and is what it is, regardless of the calculational ability. These margins compensate for aleatory and epistemic uncertainties in the determination of the risk status. We believe that any power uprate has the effect of eroding the margins. This is the reason for our recommendation that the NRC 135
l 4 staff guide its decisions on power uprates by the intent of the RG 1.174 process, which provides the appropnate rahonale forjustifying decreases in margins. Sincerely, i R.L.Seale Chairman References-1. Report dated September 17,1992, from D.A. Ward, Charman, ACRS, to James M. Taylor, Executive Drector for Operations, NRC,
Subject:
General Electnc Nuclear Energy Power Uprate Program / Fermi, Unit 2 Power increase Request.
- 2. Letter dated March 22,1996, from W. Marquino, GE Nuclear Energy, to U. S. NRC Document Control Desk, transmitting " Generic Evaluations of General Electnc Boiling -
Water Reactor Extended Power Uprate," NEDC-32523P, March 1996, and NEDC-32523P Supplement 1, Volumes I & 11, June 1996 (Propnetary). 3. GE Nuclear Energy Report NEDC-32424P, "Genenc Guidelines for General Electnc Boiling Water Reactor Extended Power Uprate," February 1995 (Propnetary).
- 4. Letter dated February 8,1996, from D. Crutchfield, NRC, to G. Gozzi, General Electric Nuclear Energy, transmitting Staff Position Concoming General Electric Boiling-Water Reactor Extended Power Uprate Program. '
- 5. Letter dated December 4,1997, from M. Hammer, Northem States Power Company, to Nuclear Regulatory Commission, transmitting Revision 1 to License Amendment Request dated July 26,1996 Supporting the Monticello Nuclear Generating Plant Power Rerate Program (includes Propnetary information).
- 6. Letter dated July 25,1997, from T. Kim, Office of Nuclear Reactor Regulation, NRC, to R.
Anderson, Northem States Power Company, transmitting Amendment Regarding Updated Analysis of DBA Containment Temperature and Pressure Response and Reliance on Containment Pressure to Compensate for Potential Deficiency in NPSH for ECCS Pumps During DBA.
- 7. Letter dated July 16,1997, from W. Hill, Northem States Power Company, to Nuclear Regulatory Commission, transmitting response to Request for Additional Information Regarding Monticello Nuclear Generating Plant License Amendment.
- 8. Letter dated July 16,1997, from W. Hill, Northem States Power Company, to Nuclear Regulatory Commission, transmitting response to Request for Additional Information Regarding Revision 2 to Monticello Nuclear Generating Plant License Amendment.
136
5 9. Memorandum dated May 18,1998, from E. Adensam, Nuclear Reactor Regulation, NRC, to J. Larkins, ACRS, transmitting Staff Position Conceming GE Licensing Topical Report NEDC-32523P on Generic Evaluation of Boiling Water Reactor Extended Power Uprate.
- 10. Memorandum dated May 14,1998, from E. Adensam, Nuclear Reactor Regulation, NRC, to J. Larkins, ACRS, transmitting Staff Safety Evaluation of the Lead Plant (Monticello)
Application.
- 11. U. S. Nuclear Regulatory Commission, SECY-98-015, " Final General Regulatory Guide ,
and Standard Review Plan for Risk-Informed Regulation of Po ver Reactors," dated January { 30,1998, transmitting Regulatory Guide 1.174, "An Approac-1 for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis." i I l l l 137
l
#' 'o g UNITED STATES / o NUCLEAR REGULATORY COMMISSION y I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Q I WASHINGTON. D. C. 20556 v
o****g September 9,1998 l MEMORANDUM TO: L Joseph Callan Executive Director'- ^ - ^n FROM: John T. Larkins, ec irector Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED FINAL RULEMAKING: CHANGES TO PARAGRAPH (h) OF 10 CFR PART 50.55a, " CODES AND STANDARDS" During the 455th meeting of the Advisory Committee on Reactor Safeguards, September 24,1998, the Committee considered the subject rulemaking and decided not to review it. The Committee has no objection to issuing the subject rulemaking for industry use. > Reference. Memorandum dated July 20,1998, from Samuel J. Collins, Office of Nuclear Reactor Regulation, to John T. Larkins, ACRS, and T. T. Martin, CRGR,
Subject:
Final Rulemaking: Changes to Paragraph (h) of 10 CFR Part 50.55a, " Codes and Standards." cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR A. Thadani, RES T. Martin, AEOD S. Aggarwal, RES l 139
/ o,, UNITED STATES l 8 o NUCLEAR REGULATORY COMMISSION 'f I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l WASHINGTON, D. C. 20555 Q ,ooo** /
September 14,1998 1 I The Honorable Shirley Ann Jackson i Chairman ! U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
1 i
SUBJECT:
EMERGENCY CORE COOLING SYSTEM STRAINER BLOCKAGE During the 455th meeting of the Advisory Committee on Reactor Safeguards, September 2-4, 1998, we continued our review of the issue involving emergency core cooling system (ECCS) strainer blockage. During several meetings since January 1993, we discussed issues related to boiling water reactor (BWR) ECCS strainer blockage, and the staff has issued numerous regulatory documents on this matter. During these meetings, we had the benefit of discussions with representatives of the NRC staff and the BWR Owners Group (BWROG). We also had the j benefit of the documents referenced. l BACKGROUND On July 28,1992, an event at Barsebeck Unit 2, a Swedish BWR, resulted in the blockage of two ECCS pump suction strainers. Subsequently, strainer blockage precursor events also occurred at U.S. plants. As a result of these events, the staff initiated analyses to estimate the l potential for losing net positive suction head for ECCS pumps at BWR facilities and used the r:sults of these analyses to identify corrective actions. On November 20,1996, the BWROG submitted topical report NEDO-32686 " Utility Resolution Guidance for ECCS Suction Strainer Blockage," also known as the URG. The topical report, which supports licensees' implementation of the guidance in Revision 2 to Regulatory Guide (RG) 1.82, " Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant l Accident," provides a methodology for establishing strainer debris loading design criteria. The staff accepted the URG, with exceptions, as documented in the Safety Evaluation Report (SER) issued on August 20,1998. 141
2 CONCLUSIONS AND RECOMMENDATIONS . The guidance provided in Revision 2 to RG 1.82, the URG, and the associated SER will l allow licensees to design and install strairners that resolve the strainer blockage issue. , e in the SER, the staff identified several areas where it did not agree with the conclusions stated in the URG. We believe that the staff's exceptions were justified. . Generic Letter (GL) 98-04 (" Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment"] contains information that could be useful to licensees in resolving the issues related to containment coatings. Resolution of these issues will allow licensees to respond completely to Bulletin 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors." . We recommend that the staff continue working with industry groups to expeditiously complete ongoing activities associated with the ECCS strainer blockage issue. Research necessary to support these activities should be given high priority. . We recommend that generic letters not be used to specify enforcement actions. DISCUSSION Since our last report to the Commission on this issue on February 26,1996, the staff has issued Bulletin 96-03 and RG 1.82, Revision 2, to resolve the safety issue regarding ECCS strainer blockage by debris resulting from a loss-of-coolant accident. In November 1996, the BWROG submitted the URG to the staff. After much correspondence and several meetings between the staff and the BWROG, the staff issued an SER on the URG. In the SER, the staff identified several areas where it disagreed with the conclusions stated in the URG. We believe that the staff's exceptions were justified. The staff concluded, however, that in general the URG provides licensees valuable guidance for evaluating the strainer blockage issue and acceptable methods for sizing strainers. The staff determined that the data obtained by the BWROG through analytical and experimental work establish an adequate basis to conclude that the URG is reasonable and/or conservative. Licensees using the URG methodology (or resolution options not accepted by the staff) need to resolve the staff's concems with the URG identified in the SER. The staff has taken actions to address the issue of foreign material that could block an ECCS or a safety-related containment spray system flow path in pressurized water reactors (PWRs), the latest being issuance of GL 98-04. We believe that the issues discussed in GL 98-04 regarding the application and maintenance of protective coatings adequately address concems about failed coatings that could cause restrictions in PWR ECCS flow paths. In GL 98-04, the staff stated that in some circumstances failure by the licensees to meet the existing requirements could warrant enforcement actions. We do not endorse the use of generic letters to specify enforcement actions. 142
3 The staff is reaching closure on the issues addressed in Bulletin 96-03, including the maintenance of coatings and the assurance of adequate net positive suction head for ECCS pumps. The staff plans to conduct audits at four to six plants to verify the adequacy of the implementation of commitments made in response to Bulletin 96-03. In parallel, the staff and industry are pursuing research in the area of protective coatings, strainer blockage at PWRs, and adequate net positive suction head for ECCS and containment spray system pumps. We expect that the staff's review of the strainer blockage issue at PWRs will be completed in a more expeditious manner than its review of the issue at BWRs. We would like to be informed of the results of these activities when they become available. We recognize that the issue of strainer blockage is complex. Licensees not only have to consider the type and amount of debris that can cause strainer blockage, but also must evaluate plant maintenance practices, including inspection of qualified coatings in the containment. The cooperation between the staff and the BWROG has established a process for resolving this issue. I Sincerely, R. L. Seale Chairman
References:
- 1. U. S. Nuclear Regulatory Commission Generic Letter 98-04: Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment, dated July 14,1998.
- 2. U. S. Nuclear Regulatory Commission Generic Letter 96-03: Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors, dated May 6,1996.
- 3. Memorandum dated August 20,1998, from Gary M. Holahan, Office of Nuclear Reactor Regulation, to John T. Larkins, ACRS,
Subject:
Transmittal of Final Safety Evaluation Report on the BWROG's Utility Resolution Guidance for ECCS Suction Strainer Blockage, NEDO-32686.
- 4. Letter dated June 18,1997, from R. L. Seale, Chairman, ACRS, to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
Proposed Generic Letter, " Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in the Containment."
- 5. Report dated February 26,1996, from T. S. Kress, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Proposed Final NRC Bulletin 96-XX,
" Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors" and an Associated Draft Revision 2 of Regulatory Guide 143
4 1.82, " Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident."
- 6. Report dated October 14,1994, from T. S. Kress, Chairman, ACRS, to Ivan Selin, Chairman, NRC,
Subject:
Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris. 144
8_
%o UNITED STATES NUCLEAR REGULATORY COMMISSION
$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON. D. C. 20656 September 15,1998 l Mr. L. Joseph Callan ' Executive Director for Opersbons U.S. Nuclear Regulatory Commission Washington DC 20555-0001
Dear Mr. Callan:
SUBJECT:
APPLICATION FOR POWER LEVEL INCREASE FOR EDWIN 1. HATCH NUCLEAR POWER PLANT UNITS I AND 2 During the 455th meeting of the Advisory Committee on Reactor Safeguards, September 2-4, 1998, we reviewed the application by the Southem Nuclear Operating Company, Inc. (SNC) for e power level increase for the Edwin 1. Hatch Nuclear Power Plant Units 1 and 2. Our Subcommittee on Thermal-Hydraulic Phenomena discussed this matter on August 27,1998. During these meetings, we had the benefit of discussions with representatives of General Electric Nuclear Energy (GENE), SNC, and the NRC staff. We also had the benefit of the documents referenced. The SNC has requested an eight percent power level increase for each unit of the Hatch plant ta permit power operation up to 2763 MWt. The current rated power for each unit is 2558 MWt (s five percent power level increase had been approved earlier under the original GENE generic power uprate program). The current request for an additional eight percent power level increase utilized the GENE generic guidelines [ Licensing Topical Report (LTR) NEDC-32424P] cnd generic evaluations [LTR NEDC-32523P] developed for the " Extended Power Uprete" program. This generic program is intended to ensure that the extended power uprate tpplication will either meet all the regulatory requirements or have only justifiable exceptions. In its application, the SNC took minor exception in the area of startup test recommendations. In our July 24,1998 report, we concurred with the GENE generic guidelines and evaluations as qualified by the staff's safety evaluation reports. The power increase for the Hatch units is accomplished by flattening the radial power profile while holding the peak bundle power constant. The core exit steam quality is increased. Additional steam flow requires small increases in feedwater flow and reactor recirculation pump speed. The additional flow is well within the capacity of the separator and dryer system. Significant increase in the reactor vessel pressure is avoeded by modifying the high-pressure ttage of the steam turbines. The reactor core power / flow map is expanded along the current
' load line limit" so that the maximum core flow limit does not exceed the pre-uprate value. A detect-and-suppress system is used by SNC to deal with boiling water reactor instability 145
2 ; concems. This system utilizes closely spaced individual local power range monitor neutron flux j detectors, combined into cells, to detect any local or global instability (Option lll of Supplement 1 to NRC Bulletin 88-07 and NRC Generic Letter 94-02). Other changes include modification of some balance-of-plant systems; recalibration of plant instrumentation, including changes in set I points; and appropriate modification of plant procedures. I The NRC provides criteria in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists in any proposed license amendment. On the basis of its evaluation, the staff concluded that the requested power level increase does not involve a significant hazards consideration. The staff found the analyses, methods, and results submitted by SNC to be acceptable. The SNC presented substantia probabilistic risk analyses and uncertainty evaluations to support a risk-informed decision process. These analyses included reviews of initiating event frequencies, equipment failure rates, operator errors, and success criteria. Qualitative arguments were used to establish that the contributions of fire, extemal events, and events under shutdown conditions to CDF are likely to be acceptably small. The effects of power uprate on the expected value of the large, early-release frequency were shown to be small. The results of the analyses showed that the most risk-significant effect of the power uprate was a change in the failure probability associated with the operator action to depressurize when inadequate high-pressure injection exists. The resulting change 'in core damage frequencies (ACDF) for Units 1 and 2 is less than 104 per year at a base CDF (for intemal events) of about 2 X 104per year. l We concur with the conci'Jsions of the staff that the requested license amendment meets all regulatory requirements for adequate protection and does not pose undue risk to the health and safety of the public. Therefore, we agree with the staff's recommendation for approval of the requested eight percent power level increase for the Edwin 1. Hatch Nuclear Plant Units 1 and 2. . Sincerely, R. L. S e ale Chairman References-
- 1. Memorandum dated August 14,1998, from H. N. Berkow, Office of Nuclear Reactor i Regulation, to J. T. Larkins, ACRS, transmitting draft Safety Evaluation Supporting Southem Nuclear Operating Company, Inc., Application for Edwin 1. Hatch Power Uprate.
- 2. Letter dated August 8,1997, from H. L. Sumner, Jr., Southem Nuclear Operating Company, Inc., to U.S. Nuclear Regulatory Commission, transmitting Edwin 1. Hatch Nuclear Plant Request for License Amendment, Extended Power Uprate Operation (contains proprietary information).
146
3
- 3. Letter dated March 9,1998, from H. L Sumner, Jr., Southem Nuclear Operating Company, ix., to U.S. Nuclear Regulatory Commission, transmitting Response to Request for Addibonal information on Extended Power Uprote License Amendment Request.
- 4. Letter dated May 6,1998, from H. L. Sumner, Jr., Southem Nuclear Operating Company, Inc., to U.S. Nuclear Regulatory Commission, transmitting Response to Request for Addnional information on Extended Power Uprate License Amendment Request.
- 5. Latter dated July 6,1998, from H. L. Sumner, Jr., Southem Nuclear Operating Company, Inc., to U.S. Nuclear Regulatory Commission, transmitting Response to Request for Additional Information on Extended Power Uprate License Amendment Request (contains proprietary information).
- 6. Letter dated July 24,1998, from R. L. Seale, Chairman, ACRS, to L. Joseph Callan, Executive Director for Operations, NRC,
Subject:
General Electric Nuclear Energy Extended Power Uprate Program and Monticello Nuclear Generating Plant Power Level increase Request.
- 7. Letter dated July 31,1998, from H. L Sumner, Jr., Southem Nuclear Operating Company, Inc., to U.S. Nuclear Regulatory Commission, transmitting Response to Request for Additional Information on Extended Power Uprate License Amendment.
- 8. Letter dated April 17,1997, from H. L. Sumner Jr., Southem Nuclear Operating Company, Inc., to U.S. Nuclear Regulatory Commission, transmitting Revised Post-LOCA Doses.
- 9. Letter dated March 22,1996, from W. Marquino, GE Nuclear Energy, to U.S. Nuclear Regulatory Commission, transmitting Generic Evaluations of General Electric Boiling i Water Reactor Extended Power Uprate, NEDC-32523P, March 1996, and NEDC-32523P Supplement 1, Volumes 1 & 11, June 1996 (Propnetary).
- 10. GE Nuclear Energy Report NEDC-32424P, " Generic Guidelines for General Electric .
Boiling Water Reactor Extended Power Uprate," February 1995 (Proprietary). i
- 11. Letter dated February 8,1996, from D. Crutchfield, NRC, to G. Sozzi, General Electric Nuclear Energy, transmitting Staff Position Concoming General Electric Boiling Water Reactor Extended Power Uprate Program.
- 12. Memorandum dated May 18,1998, from E. G. Adensam, Office of Nuclear Reactor Regulation, to John T. Larkins, ACRS, transmitting Staff Position Concoming GE Licensing Topical Report, NEDC-32523P on Generic Evaluation of Boiling Water I Reactor Extended Power Uprate.
- 13. U. S. Nuclear Regulatory Commission Bulletin 88-07, Supplement 1: Power Oscillations in Boiling Water Reactors (BWRs), dated December 30,1988.
- 14. U.S. Nuclear Regulatory Commission Generic Letter 94-02: Long-Term Solutionr and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilitie. :n ,
Boiling Water Reactors, dated July 11,1994. 147
8 k o UNITED STATES
- NUCLEAR REGULATORY COMMISSION
$ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS waswiwarow, p. c.nosse September 30,1998 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
IMPACT OF PROBABILISTIC RISK ASSESSMENT RESULTS AND INSIGHTS ON THE REGULATORY SYSTEM During the 455th and 456th meetings of the Advisory Committee on Reactor Safeguards, September 2 4 and September 30-October 2,1998, we met with representatives of the NRC staff and the Nuclear Energy Institute to discuss the issues in the Staff Requirements Memorandum (SRM) dated April 20,1998. In the SRM, the Commission requested the ACRS to identify situation-specific cases where probabilistic risk assessment (PRA) results and insights have improved the existing regulatory system and specific areas where PRA can have 13 positive impact on the regulatory system. Our Subcommittee on Reliability and Probabilistic Risk Assessment met on August 26,1998, to discuss these issues. We also had the benefit of the documents referenced. General Observations in the past the NRC has utilized qualitative evaluations of risk, based on engineering judgment etnd experience, to carry out its mission to protect public health and safety. Uncertainties in the qualitative evaluations of risk have been addressed by adopting greater conservatism. Such qualitative evaluations of risk do not permit the most effective allocation of resources by licensees or the NRC. The traditional engineering analyses do not permit the examination of complex engineering systems like nuclear power plants in a detailed, integrated manner. Instead, safety analyses have been performed out of necessity on a subsystem-by-subsystem or even on a component-by-component basis. Interactions among systems and the unanticipated responses of multiple systems to unexpected situations can result in higher than cxpected risk even when each system or component meets all the regulatory requirements. Examples include the interfacing-systems loss-of-coolant accident that was identified by the Reactor Safety Study (WASH-1400) and the recent accident sequence, initiated by fire, identified by the Individual Plant Examination of Extemal Events (IPEEE) for the Quad Cities Nuclear Power Plant. The NRC has recognized the limitations of traditional engineering analyses and has pioneered the development of the quantitative risk assessment of nuclear power plants, namely, PRA 149
2 methodology. PRA builds upon traditional engineering analyses to develop quantitative assessments of risk. In fact, it is the only means for quantitative determination of risk. PRA methodology examines safety systems and their interactions in an integrated, comprehensive manner. It is scrutable in that engineering judgments are quantified. It yields quantitative measures of risk significance for individual systems, structures, and components (SSCs) that can provide a basis for a more efficient allocation of resources by Heensees and the NRC. Although uncertainties are present in any type of analysis, many of the uncertainties can be quantified in PRA and this quantification should be used to constrain conservatism. Risk to the I public can be expressed in terms comparable with other risks and objective definitions can be developed for the NRC's mission goal of preventing " undue risk." The principal strengths of the current regulatory system are its caution ("what if we are wrongT) and the resulting development of the principle of defense-in-depth as well as its large experience base. The principal weakness is its inability to quantify the risk significance of SSCs. The principal strengths of PRA are the quantification of risk and the identification and ranking of the major accident sequences and risk-significant SSCs. The principal weakness is incompleteness, i.e., the inability to identify all potential threats to the system and to develop i adequate models for some identified threats. The principle of defense-in-depth and PRA can complement each other well. When the uncertainties in the PRA are too large for regulatory decisionmaking (especially due to incompleteness), the principle of defense-in-depth can be invoked to deal with these uncertainties. Although this may appear obvious, it should be emphasized because it is too often ignored. The strengths and limitations of both the current system and PRA must be evaluated when a new regulatory application is contemplated. We believe that combining the strengths of PRA and defense-in-depth will lead to better-informed decisionmaking and improved regulatory coherence. We anticipate that it will also lead to greater regulatory efficiency and reduction of unnecessary burden on both licensees and the NRC staff. Past and Current improvements in the Regulatory System Some specific examples where PRA has improved (or is expected to improve) the current regulatory system are:
- 1. The Anticipated Transients Without Scram (ATWS) Rule. PRA identified the importance of ATWS and provided the technical basis for regulatory action.
- 2. The Station Blackout (SBO) Rule. PRA identified the significance of SBO and provided the technical basis for regulatory action. PRA permitted the assessment of the risk from SBO on the basis of plant-specific configuration, as well as plant-specific grid, switchyard, and weather characteristics. It permitted evaluation of the cost effectiveness of attemative improvements, thereby leading to more efficient allocation of resources. This illustrates one of the strengths that PRA brings to the regulatory process.
i 150
l 3 j
- 3. Generic Safety Issue Prioritization and Resolution. Using PRA criteria, the original list of \
about 700 generic safety issues was reduced to about 200, thereby focusing NRC resources on the most important issues.
- 4. Advanced Reactor Design Certifications (ABWR., AP600, CE System 80+). PRA allowed the staff and each vendor to focus on the design issues important to safety, thereby leading to substantial reduction in risk for these designs.
- 5. Licensing Amendments. The recently issued regulatory guides allow the use of risk information in requests for changes in the technical specifications, inservice testing, and quality assurance requirements (the inservice inspection regulatory guide is still under consideration). The principal benefits are expected to be improved safety and efficient j allocation of resources. Graded quality assurance provides an example where, even though the impact of quality assurance requirements on PRA is unquantified, one can still derive insights regarding the importance of SSCs from PRAs.
Future improvements of the Regulatory System
- 1. Oversight Process. There is a widespread belief within the industry that the current inspection and enforcement processes are overly prescriptive and burdensome. Plant-specific risk information can and should be used to focus regulatory and licensee attention. Enforcement actions, too, should be graded in terms of risk-significance.
- 2. 10 CFR 50.59 Process. The strength of the current process is that it ensures that l changes made without prior NRC approval do not constitute an unreviewed safety question in accordance with the Final Safety Analysis Report (FSAR) which is the basis for licensing the facility. The major weakness is that the process refers to changes in probabilities that cannot be calculated using traditional deterministic methods. 1 Furthermore, the Commission recently directed the staff to define " minimal" changes to ensure that such changes are sufficiently small that NRC review is not required. The quantification of frequencies of events, one of the strengths of PRA, provides the context within which contemplated changes can be declared " minimal." At the same ,
time, one of the weaknesses of PRA is its insensitivity to very small changes in plant ! I configuration and procedures. We, therefore, expect that a revised 10 CFR 50.59 process will retain parts of the " deterministic" criteria that the current process employs. l
- 3. Fire Protection. The recent discovery of a major accident sequence, initiated by fire, identified by the IPEEE at Quad Cities demonstrated the limitations of the existing fire protection regulations. A revision of 10 CFR 50, Appendix R to include risk information would reduce the likelihood that such cases would reoccur. Such a revision would, of course, have to take into account the limitations of current fire risk assessment methodology (e.g., the lack of models for assessing the impact of smoke) and would rely on defense-in-depth. PRA would also be useful in the prioritization of inspections of fire barrier penetration seals that have been of concern recently, thereby avoiding the waste of resources on insignificant issues.
151
4- !
- 4. Prioritization of Research Needs. In an era of diminishing budgets, it is no longer sufficient to rely primarily on judgment to prioritize research. The principal criterion for prioritizing research needs should be their expected contribution to risk-infonrned regulatory decisions.
- 5. Assessment of Changes in Post-Three Mile Island Requirements. Many requirements were imposed in the immediate aftermath of the accident at Three Mile Island Unit 2.
These changes did not have the benefit of significant input from PRA, which was a developing technology at the time. The risk importance of these requirements should be evaluated. Based on these evaluations, the requirements may be changed or eliminated. : Transition to Risk-informed Regulation The transition to a risk-informed regulatory framework will be incremental. Many of the present regulations are based on deterministic and prescriptive requirements that cannot be quickly replaced. Therefore, the current requirements will have to be maintained while risk-informed regulations are being developed and implemented. Furthermore, we expect that a number of licensees will, for a variety of reasons, be unwilling to embrace a new regulatory system. Therefore, the NRC should be prepared to accommodate a two-tier system, i.e., a modified version of the current regulatory process and a risk-informed system. This situation will prevail for a number of years and may create circumstances that should be addressed by the Commission. We have already seen such circumstances in recent requests for BWR power uprates. Even though licensee use of Regulatory Guide 1.174 is voluntary, questions were raised about the acceptability of the change in core damage frequency associated with power uprates. Although in this case the licensees voluntarily submitted the relevant information, conflicts might arise in the future. Although we recognize that it will be necessary to maintain many of the current requirements during the transition, we strongly support the efforts of the staff to develop options to revise 10 CFR Part 50 to make it more risk informed. We believe that, as a minimum, revisions must be made to permit effective implementation of the initiatives associated with Regulatory Guide 1.174. l An example of the need for regulatory harmonization is the attempt to apply the recently issued , Regulatory Guide 1.176 on Graded Quality Assurance. This Regulatory Guide utilizes PRA importance measures to categorize SSCs according to their safety significance. Industry representatives have stated that they expect that several thousand components, which are currently classified as safety-related, will be placed in the " low-safety significance" category, which indicates that quality assurance requirements on these components could be relaxed with little impact on safety. It is not clear whether, under the current regulations, this relaxation of requirements can be done under 10 CFR 50.59 or whether each request must be submitted to the staff for review and approval, in which case the potential benefits of graded quality assurance will be reduced significantly. We anticipate that similar cases will arise in the future. To further the use of PRA in the regulatory process, we recommend that the Commission consider some policy decisions. First, determine whether risk itself or surrogate measures such 152
1 5 Os core damage frequency are to be used in making decisions based on PRA. Second, direct the staff to allow credit for voluntary actions consistent with the Commission directive that risk Casessments be as realistic as possible. Finally, we recommend that the Commission expedite the revision of 10 CFR 50.12 to allow the use of risk insights as a basis for exemptions to its current regulations. The development of PRA technology should be continued. For example, a good understanding of risk is needed in the following areas: low-power and shutdown operations, fire protection systems, software-based digital systems, and measures of safety culture. Sincerely, R. L. Seale Chairman References-1 Memorandum dated April 20,1998, from John C. Hoyle, Secretary, NRC, to John T. Larkins, Executive Director, ACRS,
Subject:
Staff Requirements - Meeting with the Advisory Committee on Reactor Safeguards, April 2,1998.
- 2. Memorandum dated August 18,1998, from John C. Hoyle, Secretary, NRC, to L.
Joseph Callan, Executive Director for Operations, NRC, and Karen D. Cyr, General Counsel, NRC,
Subject:
Staff Requirements - Public Meeting on Stakeholder Concems, July 17,1998.
- 3. Memorandum dated August 7,1998, from Shirley Ann Jackson, Chairman, NRC, to L.
Joseph Callan, Executive Director for Operations, NRC,
Subject:
Responding to issues Raised within the Senate Authorization Context.
- 4. Memorandum dated August 25,1998, from L. Joseph Callan, Executive Director for Operations, NRC, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Response to issues Raised within the Senate Authorization Context and July 17,1998 Stakeholder Meeting. 153
'o,, UNITED STATES 8 ,,
NUCLEAR REGULATORY COMMISSION 'J
- I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g / wAsumcTos.o.c. zoons
'% or ,*** F October 6,1998 MEMORANDUM TO: L. Joseph Callan -
Executive Director for 0;erations o - FROM: John T. Larkins, Executive Director *
')
Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED FINAL AMENDMENT TO 10 CFR PART 55, "lNITIAL LICENSED OPERATOR EXAMINATION REQUIREMENTS" During the 456th meeting of the Advisory Committee on Reactor Safeguards, September 30 - October 2,1998, the Committee considered the. proposed final amendment to l 10 CFR Part 55 and decided not to review it. The proposed changes do not alter the l Committee's previous recommendation on this matter. The Committee has no objection to 1 issuing this amendment for industry use.
Reference:
l Memorandum dated September 22,1998, from R. Lee Spessard, NRR, to Addressees,
Subject:
Office Review and concurrence on Final Rule - Requirements for Initial Operator Licensing Examinations - Amendment to 10 CFR Part 55. cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR R. Spessard, NRR 155
/ [ o,, o UNITED STATES NUCLEAR REGULATORY COMMISSION fo I ADVISORY COMMITTEE ON REACTOR SAFE!UARDS wAsHwaTow. o. c. rosss
.o...
October 6,1998 MEMORANDUM TO: L. Joseph Callan Executive Director for Operations - FROM: John T. Larkins, Executive Director , Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED RULE--" CHANGES TO REQUIREMENTS FOR i ENVIRONMENTAL REVIEW FOR RENEWAL OF NUCLEAR POWER PLANT OPERATING LICENSES (10 CFR PART 51)" During the 456th meeting of the Advisory Committee on Reactor Safeguards, September 30 - October 2,1998, the Committee considered the subject rule and decided not ta review it. The Committee has no objection to issuing the proposed rule for public comment. Reference Memorandum dated September 14,1998, from Jack W. Roe, NRR, to Addressees,
Subject:
Proposed Rule" Changes to Requirements for Environmental Review for Renewal of Nuclear Power Plant Operating Licenses (10 CFR Part 51)." cc: J. Hoyle, SECY J. Blaha, OEDO ' J. Mitchell, OEDO S. Collins, NRR J. Roe, NRR D. Cleary, NRR J. Gray, OGC 157
r
#g UNITED STATES 8 NUCLEAR REGULATORY COMMISSION $ p} ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g wAsmwarom, p. e.zonsa October 7,1998 l
MEMORANDUM TO: L. Joseph Callan Executive Drector for Operations p FROM: John T. Larkins, Executive Drector Advisory Committee on Reactor Safeguards
SUBJECT:
PROPOSED FINAL REVISIONS TO REGULATORY GUIDE (RG) 1.84 (DESIGN AND FABRICATION CODE CASE ACCEPTABILITY), RG 1.85 (MATERIALS CODE CASE ACCEPTABILITY, AND RG 1.147 (INSERVICE INSPECTION CODE CASE ACCEPTABILITY) During the 456th meeting of the Advisory Committee on Reactor Safeguards, September 30 - October 2,1998, the Committee considered the proposed final revisions to the subject regulatory guides and decided not to review them. The Committee has no objection to issuing these revisions for industry use. Reference-Memorandum dated September 16,1998, from Lawrence C. Shao, RES, to John T. Larkins, Executive Director, ACRS,
Subject:
Request for ACRS Rev'mw and Comment of Regulatory Guide 1.84 (Design and Fabrication Code Case Acceptability, ASME Section Ill, Division 1), Regulatory Guide 1.85, (Materials Code Case Acceptability, ASME Section lil, Division 1), and Regulatory Guide 1.147 (Inservice inspection Code Case Acceptability, Section XI, Division 1). cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO A. Thadani, RES L. Shao, RES M. Mayfield, RES 159
#o, UNITED STATES f r, NUCLEAR REGULATORY COMMISSION 'fo I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHWGTON, D. C. 20556 October 8,1998 MEMORANDUM TO: L. Joseph Callan Executive Director for Operations FROM: John T. Larkins, Executive Director a u-<--O Advisory Committee on Reactor Safeguards -
SUBJECT:
PROPOSED FINAL REGULATORY GUIDE DG-1029, REVISION 1.7," GUIDELINES FOR EVALUATING I ELECTROMAGNETIC AND RADIO-FREQUENCY INTERFERENCE IN SAFETY-RELATED INSTRUMENTATION AND CONTROL SYSTEMS" During the 456th meeting of the Advisory Committee on Reactor Safeguards, September 30 - October 2,1998, the Committee considered the subject Regulatory Guide and decided not to review it. The Committee has no objection to issuing the proposed final regulstory guide for industry use. Reference Memorandum dated September 23,1998, from Thomas L. King, Office of Nuclear Reactor Regulation, to John T. Larkins, ACRS,
Subject:
Proposed Final Regulatory Guide on Electromagnetic Interference and Radio-Frequency Interference i cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO A. Thadani, RES C. Antonescu, RES 161
/ 'g UNITED STATES
/ NUCLEAR REGULATORY COMMISSION y ADVISORY COMMITTEE ON REACTOR SAFEGUARDS e WASHINGTON, D. C. 20056 October 14,1998
)
- i Mr. L. Joseph Callan Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Mr. Callan:
SUBJECT:
RISK-INFORMED PILOT APPLICATION FOR HYDROGEN MONITORING AT ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 During the 456th meeting of the Advisory Committee on Reactor Safeguards, September 30 - October 2,1998, we reviewed the risk-informed pilot application for monitoring hydrogen concentration in containment at Arkansas Nuclear One (ANO), Units 1 and 2. In this application, the licensee requested that the required time for activating the hydrogen monitoring system after start of safety injection be changed from 30 minutes to 90 minutes to reduce i burdens on operators at critical times. During this review, we had the benefit of discussions with representatives of the NRC staff and with a representative of Performance Technology, , Inc. We also had the benefit of the documents referenced. Recommendation We agrt? ,vith the supporting analyses for the ANO licensee's request and have no objection to the staff's approval.
)
l Discussion l Entergy Operations, Inc. (EOl) requested relief from the requirement that the hydrogen monitoring system be activated within 30 minutes following the start of safety injection. EOl stated that the need for monitoring the hydrogen concentration for design-basis accidents (and presumably for higher probability accidents) only occurs after several hours following safety initiation. They also demonstrated that the hydrogen recombiners have insufficient capacity to cignificantly mitigate tne hydrogen concentration resulting from severe accidents. Any short-term need to have early indication of @ure damage status is satisfied by other more appropriate and usefulindicators. The first 30 minutes after the start of safety injection is a crucial period in which plant operators cre called upon to take numerous high-priority actions. The requirement to activate the hydrogen monitoring system during this period is an unnecessary diversion. EOl made a persuasive qualitative case that the removal of the diversion with this requested change has a high likelihood of actually decreasing risk. Inasmuch as defense-in-depth and the deterministic 163
regulatory requirements also appear to be appropriately treated in this change request, we believe that it would qualify as being acceptable under the Regulatory Guide (RG) 1.174 guidance. Although the licensee did not elect to use this approach, we believe that RG 1.174 provides appropriate guidance for the staff's review. Although it is apparent that this requested change does not pose any undue risk, other, more significant, changes to the hydrogen recombiner systems could have implications with respect to the ability to manage or limit releases of smaller quantities of fission products from unfailed containments. The value of recombiner systems in this regard should be quantified prior to making decisions on licensee requests for removal of, or other significant changes to, these systems. Sincerely, R. L. Seale Chairman References
- 1. Letter (undated), W. Reckley, NRR, to C.R. Hutchinson, Entergy Operations ANO, transmitting Confirmatory Order Modifying Post-TMl Requirements Pertaining to Containment Hydrogen Monitors for Arkansas Nuclear One, Units 1 and 2 (Predecisional Draft).
- 2. Letter dated September 9,1998, from J. D. Vandergrift, Entergy Operations, Inc., to NRC,
Subject:
Proposed Change to Requirements Regarding Containment Hydrogen Monitors for Arkansas Nuclear One, Units 1 & 2.
- 3. Letter dated March 2,1998, from D.C. Mims, Entergy Operations Inc., to NRC,
Subject:
NUREG-0737, item ll.F.1.6, Containment Hydrogen Monitor Request for Relief.
- 4. Letter dated October 28,1997, from D.G. Mcdonald, Office of Nuclear Reactor Regulation, NRC, to N. S. Cams, Northeast Nuclear Energy Company,
Subject:
Withdrawal of Deviation Request for NUREG-0737, item II.F.1.6, Containment Hydrogen Monitors - Millstone Nuclear Power Station, Unit No. 2.
- 5. Letter dated February 4,1992, from J. J. Fisicaro, Entergy Operations, Inc., to NRC,
Subject:
NUREG-0737, item II.F.1 Attachment 6, Hydrogen Analysis Capability.
- 6. Letter dated September 11,1998, from, B. Christie, Performance Technology, to R. L.
Seale, Advisory Committee on Reactor Safeguards, Chairman, transmitting information relevant to " Risk-informed Pilot Application for Hydrogen Monitoring at Arkansas W: lear One."
- i. tE.ter dated September 21,1998, from B. Christie, Performance Technology, to R. L. j Seale, Advisory Commi+be o') Reactor Safeguards, Chairman, Transmitting Letter dated September 10,1998, from D. E. Nunn, Southem Califomia Edison,
Subject:
Request for Exemption to 10CFR50.44,10CFR50, Appendix A, General Design Criterion 41, and 10CFR50, Appendix E, Section VI. Proposed Technical Specification Change NPF-10/15-496, San Onofre Nuclear Generating Station, Units 2 and 3 , (SONGS 2 & 3) ! 164
@ Usu o,, UNITED STATES 8 n NUCLEAR REGULATORY COMMISSION $ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g, wAssmoTos, p. c.zosss October 16,1998 Mr. L. Joseph Callan Executive Director for Operations U. S. Nuclear Regulatory Commission Wrshington, D.C. 20555-0001 De:r Mr. Callan:
SUBJECT:
PROPOSED PRIORITY RANKINGS OF GENERIC SAFETY ISSUES: TENTH GROUP During the 455th and 456th meetings of the Advisory Committee on Reactor Safeguards, September 2-4 and September 30-October 2,1998, we reviewed the priority rankings proposed by the NRC staff for the Generic Safety issues (GSis) listed in Table A. During our review, we had the benefit of discussions with representatives of the NRC staff. We also had the benefit of the documents referenced. Our comments on various GSis considered during these meetings are contained in the following attachments: Attachment 1 lists those GSis for which we agree with the priority rankings proposed by the NRC staff. Attachment 2 identifies the GSis for which we agree with the priority rankings proposed by the staff, but have comments. Attachment 3 identifies the GSI for which we disagree with the priority ranking proposed by the staff. } In cddition to the comments on the proposed prionty rankings, we offer the following comments end recommendations on the GSI process:
- Overall GSI Process In recent years, the GSI process has not functioned property. This may be attributed to frequent changes in management responsible for its implementation. The staff should improve its capability to perform cost / benefit analysis and to use the risk-informed approach in prioritizing and resolving GSis.
165
2 j i
- Pnontszation in SECY-98166," Summary of Activities Related to Generic Safety lasues," the staff proposes to discontinue use of the term "nearly resolved" and to revise NUREG-0933 to reflect this change in terminology. This proposed action will resolve our concem regarding the use of the term *nearly resolved," which has been a misleading category in the GSI prioritization process For example, GSI-190, " Fatigue Evaluation of Metal Components for 60-Year Plant'Ufe," which was classified as "Nearly Resolved" in 1996, is yet to be resolved. Similarly, GSI-191," Assessment of Debns Accumulation on PWR Sump Performance," was categorized as "Nearty Resolved"in 1996. In the document provided to us, the staff states that research will be initiated in FY 1998 to resolve this issue, which may take several years to complete.
Several of the 20 GSIs provided for our review were categorized as "Nearly Resolved." In SECY-98-166, however, these issues were arbitrarily recategonzed as HIGH. The basis for these rankings should be documented. Another category used in the pdeisetion process is " Resolved " We believe that there have been cases where this term is used too loosely. For example, the existence of a plan to resolve a particular GSI does not necessarily mean that the issue has been technically resolved. The staff should ensure that an adequate technical basis exists prior to declaring that a GSI has been resolved. As part of a reevaluation of the GSI process, thought should be given to the appropnateness of using the classifications " Regulatory impact Issue,"" Licensing issue," and " Environmental issue"in the prioritization process, irrespective of additional terminology applied to an issue, we believe that all issues should be priontized as HIGH, MEDIUM, LOW, or DROP to provide consistency throughout the Agency. The
. emphasis by the Commission on reducing unnecessary regulatory burden to the industry supports the need for prioritizing the issues placed under the above three categories. Also, resolution of these issues should be clearly documented.
In our March 16,1998 report, we noted that the planning by the Office of Nuclear Regulatory Research calls for the prioritization of two to three GSis per year. In our April 12,1988 report to the Commission, we stated that the average time required to assign a priority to a GSI is about six months, which we do not consider unreasonable. It is disturbing to see that the range of times involved in prioritizing GSis varies widely. For example, GSI-163, " Multiple Steam Generator Tube Leakage," identified in 1992, was not prioritized until 1997, and GSI-169, "BWR MSIV Common Mode Failure Due to Loss of Accumulator Pressure," identified in 1993, was not prioritized until 1998. The staff should take effotts to ensure that GSis are prioritized expeditiously. The methodology used in the prioritization process is technically sound, but the staff should ensure the quality and appropriateness of the assumptions used in the analysis supporting the priority ranking of a particular GSI. For example, we recently reviewed the proposed resolution of GI-171, "ESF Failure From LOOP Subsequent to a LOCA," which was assigned HIGH pnonty ranking in 1995. Based on reassessment of the 1 166 l l
I l assumptions and the frequency numbers used in calculating the core damage frequency ; (CDF), the CDF decreased by three orders of magnitude . This raises concem about ' the veldity of the assumphons and analyses used in pnontizing other GSis.
- Resolution Fifteen of the GSis identified since the 1979 amendment to the Energy Reorganization Act of 1974 have still not been resolved. We strongly urge contmuod effort to resolve these issues 1 The staff has assumed that the safety concems associated with several GSis would be addressed by the licensees in the individual plant examination / individual plant examination for extemal events (IPE/IPEEE) programs. We recommend that after completing the review of the IPE/IPEEE submittals, the staff provide a report documenting whether the conooms of these GSis were, in fact, addressed adequately so that they can be considered resolved. Those issues that were not adequately addressed should be pnontized and resolved.
- Coordinahon The senior management of the Office of Nuclear Reactor Regulation and the Office of Nuclear Regulatory Research should ensure adequate coordination between their offices to resolve technical differences associated with GSls in a timely manner to facilitate expeditious piio iis. tion and resolution of GSis.
Sincerely, R. L Seale Chairman Attachments: As stated . References-
- 1. Memorandum dated July 6,1998, from L Joseph Callan, Execuhve Director for Operations, NRC, for The Commissioners,
Subject:
SECY-98-166, " Summary of i Activities Related to Generic Safety issues."
- 2. Report dated March 16,1998, to L. Joseph Callan, Executive Director for Operations, NRC, from R. L Seale, Chairman, ACRS,
Subject:
SECY-98-001, Mechanism for Addressing Generic Safety issues. , i
- 3. Report dated April 12,1988, to the Honorable Lando W. Zech, Jr., Chairman, NRC, from W. Kerr, Chairman, ACRS,
Subject:
Effectiveness of Programs Relating to Generic and Unresolved Safety issues - ACRS Comments. Letters dated February 24,1998, to The Honorable Albert Gore, Jr., President of the United States Senate, and The Honorable Newt Gingrich, Speaker of the United States 167
House of Representatives, from R. L Seale, Chairman, ACRS, transmitting " Nuclear Safety Research, A Report to the U.S. House of Representatives and the U.S. Senate."
- 5. Memorandum dated September 16,1993, to James M. Taylor, Executive Director for Operations, NRC, from J. Emest Wilkins, Jr., Chairman, ACRS,
Subject:
Proposed Priority Rankings of Generic issues: Eighth Group. I
)
l i 1 168
I
-S-TABLE A l TENTH GROUP OF GENERIC SAFETY ISSUES REVIEWED BY THE ACRS DURING THE 455TH MEETING. SEPTEMBER 2-4 1998 i
Generic Safety issue Number Title Priority Ranking Proposed by the NRC Staff 163 Multiple Steam Generator Tube HIGH I Leakage 169 BWR MSIV Common-Mode DROP Failure Due to Loss of (Based on the impact /value ratio and the Accumulator Pressure total risk reduction potential, this issue is in the drop category.) 170 Fuel Damage Criteria for High HIGH Bumup Fuel (Current data cannot be correlated to design criteria and conclusive data will not be available for several years. Research is continuing on assessing the adequacy of fuel damage criteria at high bumups.) i 172 Multiple System Responses HIGH Program (Data are being collected to evaluate the mannerin which the MSRP concems were addressed by licensees in their IPE/IPEEE submittals. Staff assessment of licensee submittals will determine whether the concems have been adequately addressed.) 169
Generic Safety issue Number Title Priority Ranking Proposed by the NRC ; Staff 173 A Spent Fuel Storage Pool for HIGH l Operating Facilities (The staff is in the process of revising its J guidance documents for spent fuel storage design (i.e., portions of SRP ; 9.1.3 and Regulatory Guide 1.13). Currently, the stanis working with industry (an ANS Subcommittee) to revise ANSl/ANS-57.2, the standard that contains guidance for spent fuel storage pool design. The staff plans to incorporate the improvements from this standard into a revised SRP and Regulatory Guide. The expected completion date for issuance of the revised guidance documents is August 2000.) 173 B Spent Fuel Storage Pool for HIGH (Resolved) Permanently Shutdown Facilities (No generic action was required.) 174 A' Fastener Gaging Practices RESOLVED (This issue was resolved and no new requirements were established.) 174 B Johnson Gage Company Concem RESOLVED (This issue was resolved and no new requirements were established.) 175 Nuclear Power Plant Shift Staffing RESOLVED (This issue was resolved and no new requirements were established.) 176 Loss of Fill-Oilin Rosemount RESOLVED Transmitters (This issue was resolved and no new requirements were established.) 177 Vehicle intrusion at TMI RESOLVED j (This issue was resolved and no new requirements were established.) 178 E#ect of Hurricane Andrew on RESOLVED Turkey Point 170
Generic Safety issue Number Title Priority Ranking Proposed by the NRC Staff 179 Core Performance LICENSING ISSUE (Resolved) (This issue addresses the staWs efforts in clarifying existing requirements and guidance and, therefore, is classified as a Licensing issue. This issue was resoived with the issuance of the revised staff guidance.) 180 Notice of Enforcement Discretion RESOLVED i (This issue was resolved with the issuance of the revised staff guidance.) 181 Fire Protection LICENSING ISSUE (This issue addresses the stafs efforts in improving its capability to make independent assessments of safety and is classified as a Licensing issue. NRR is in the process of completing pilot Fire Protection Functional inspections.) 182 General Electric Extended Power REGULATORY IMPACT ISSUE Uprate (This issue does not affect safety but could have an economic impact on the operation of plants with GE reactors. Therefore,it was classified as a Regulatory impact issue.) 183 Cycle-Specific Parameter Limits in RESOLVED Technical Specifications 184 Endangered Species ENVIRONMENTAL ISSUE (This issue addresses impact on the environment of nuclear plants and, therefore, is classif'ed as an Environmental issue.) l 1 I i l 1 71
Generic Safety issue Number Title Priority Ranking Proposed by the NRC , Staff ; 190 Fatigue Evaluation of Metal HIGH Components for 60-Year Plant (The staff is studying the risk of failure Life from fatigue of selected components. A report," Fatigue Analysis of Components for 60-Year Plant Life"is under way, l making use of updated fatigue design curves for stainess steel developed by Argonne National Laboratory in March 1998. This issue is expected to be resolved by March 1999.)
]
191 Assessment of Debris HIGH Accumulation on PWR Sump (Research is being planned on coatings Performance and debris transport to determine the potential severity of PWR sump blockage effects. This work will be initiated in FY 1998 and may take several years to complete.) i i 172
( f -7ACHMENT 1 UST OF GU "EHlO SAFETY ISSUES FOR WHICH THE ACRS AGREES WITH THE PRIORITY RANKINGS PROPOSED BY THE NRC STAFF Genenc Safety lasue No. Iglg 163 Multiple Steam Generator Tube Leakage 16g BWR MSIV Common-Mode Failure Due to Loss of Accumulator Pressure 172 Multiple System Responses Program lasue 3 Failure Modes of Digital Computer Control Systems issue 4 Specific Scenarios Not Considered in USI A-47 issue 5 Effects of Degradation of HVAC Equipment on Control and Protection Systems Issue 6 Failure Modes Resulting From Degraded Electnc Power Sources l
/ssue 7 Failure Modes Resulting From Degraded Compressed Air Systems
/ssue 8 Potential Effects of Untimely Component Operation lasue 9 Propagation of Environments Associated Wdh DBAs lasue 11 Synergistic Effects of Harsh Environmental Conditions issue 12 Environmental Qualification of Seals, Gaskets, Packing, and Lubncating Fluids Associated With Mechanical Equipment 173 A Spent Fuel Storage Pool for Operating Facilities 173B Spent Fuel Storage Pool for Permanently Shutdown Facilities 174 A Fastener Gaging Practices - SONG's Employees' Concem 174 B Fastener Gaging Practices - Johnson Gage Company Concem 176 Loss of Fill-Oil in Rosemount Transmitters 173
177 Vehicle intrusion at TMI 178 Effect of Hunicane Andrew on Turkey Point 179 Core Performance 180 Notice of Enforcement Discretion 181 Fire Fmei 182 General Electric Extended Power Uprate 184 Endangered Speces 190 Fatigue Evaluation of Metal Components for 60-Year Plant Life 191 Assessment of Debris Accumulation on PWR Sump Performance 1 74
ATTACHMENT 2 LIST OF GENERIC SAFETY ISSUES FOR WHICH THE ACRS AGRFFS WITH THE PRIORITY RANKINGS PROPOSED BY THE NRC STAFF. BUT WITH COMMENTS Generic Safety lasue No. : 170 Tilm: Fuel Damage Critens for High Bumup Fuel Pnonty Ranking HIGH Prooosed by the NRC Staff : ACRS Comments : The research program that will technically resolve this issue is directed toward providing confirmatory evidence in support of regulatory deemions that have been made, The research program should ensure that adequate technical foundations and analytical tools are available to the NRC line organizations to meet regulatory needs. The research program needs to resolve criticisms leveled by NRC contractors concoming the adequacy of the treatment of delayed neutron fraction in neutron transport codes. The research program needs to document peer review arguments that criticality events will not occur if fuel is dispersed in fuel channels by credible reactivity insertion events. There must be confidence that local fuel damage does not propagate into large regions. The research program needs to ensure that Baker-Just clad exidation kinetics used in Appendix K analyses are bounding for high-bumup fuel whose clad is susceptible to thermal stress fracture and breakaway oxidation. The research program also needs to develop plans to examine high bumup fuel behavior during anticipated transients without scram (ATWS) events and ATWS recovery processes. Generic Safety lasue No: 172 Igla. Multiple System Responses Program Prionty Ranking HIGH Proposed by the NRC Staff. ACRS Comments: 1 Of the 21 Multiple System Responses Program (MSRP) issues,11 issues were to be I addressed in the IPE/IPEEE programs. After reviewing the IPE/IPEEE submittals by the ; licensees, the staff plans to prepare a summary report on how these 11 issues were addressed in the IPE/IPEEE programs. In the summary report, the staff should document clearly whether these issues have been adequately addressed by the licensees in the IPE/IPEEE programs. 175
Those issues found to be not addressed properly should be reprioritized and resolved expeditiously. Subsequent to reviewing the staff's summary report, we will decide on the adequacy of the treatment of these 11 issues in the IPE/IPEEE programs. Issue 10: Evaluation of Heat, Smoke, and Water Propagation Effects Resulting From Fires This issue addresses the question about how effluents and heat generated during a fire might disperse from the site of the fire and affect equipment in other locations. The staff plans to address the effects of environmental stressors on digital electronic equipment, including the effects of smoke as a separate issue. We plan to review the proposed resolution of this issue. Genenc Safety lasue No. 175 h Nuclear Power Plant Shift Staffing Pnonty Ranking RESOLVED Proposed by the NRC Staff. ACRS Comments-The staff should continue to monitor operating events and incidents to provide feedback regarding operational challenges and reassess the adequacy of staffing and task allocation, as appropriate. Genenc Safety Issue No : 183 Itla: Cycle-Specific Parameter Limits in Technical Specifications Priority Rankino RESOLVED Proposed by the NRC Staff: ACRS Comments : Performance by an individual licensee should not be used as the basis for closure of generic safety issues that are intended to reduce the regulatory burden on the nuclear industry. We recommend that the regulatory requirements identified by the Regulatory Review Group as being candidates for elimination be reconsidered under the generic safety issue process. 1 J 176
~
ATTACHMENT 3 GENERIC SAFETY ISSUE FOR WHICH THE ACRS DISAGREES WITH THE PRIORIT( RANKING PROPOSED BY THE NRC STAFF Genanc Safety lasue No.: 172 Ilda. Multiple System Responses Program l lasue 16: Seismically Induced Fires Priority Ranking DROP Proposed by the NRC Staff ACRS Comments: l Seismic events can cause fires and, at the same time, damage the capacity to suppress fires because fire suppression systems are not adequately qualified for sesmic events. The staff acknowledges the existence of the issue and expects that it will be adequately addressed in the IPEEE process. At the same time, the staff har. 3dentified some 12 major issues with the l industry-developed tool, Fire-induced Vulnerability Evaluation (FIVE) Methodology, for analysis l of fire and some 42 deficiencies of probabilistic risk assessment techniques for the analysis of l fire. It would seem unlikely that even the most diligent licensee efforts to address the issue in l its IPEEE program would yield persuasive results. It seems that the issue must remain open until we have a chance to review the findings of the IPEEE effort. l l 177 L__ _ _ _ - - - - - - -- - - - -
r
/ %,, UNITED STATES 8 NUCLEAR REGULATORY COMMISSION $ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wasHmorow, p.c.sonas October 20,1998 l
l The Honorable Shirley Ann Jackson l Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Dear Chairman Jackson-l
SUBJECT:
THE NUCLEAR ENERGY INSTITUTE'S PETITION FOR RULEMAKING TO l AMEND PARAGRAPH (a) OF 10 CFR 50.54, CONDITIONS OF LICENSES During the 456th meeting of the Advisory Committee on Reactor Safeguards, September 30-October 2,1998, we reviewed a draft Commission paper which documents the proposed NRC , staff position on the Nuclear Energy Institute's (NEl's) petition for rulemaking to amend 10 CFR t 50.54 (a). We also heard presentations by and held discussions with representatives of the l NRC staff and the NEl concoming such a petition which was submitted in 1995 by NEl. The l petition sought to broaden the scope of allowed unilateral changes that would not require prior l NRC review and approval. They proposed to amend 10 CFR 50.54(a) to make changes j cxempt if they do not involve an unreviewed safety question as defined in 10 CFR 50.59. We ciso had the benefit of the documents referenced. In its response, the staff agreed that 10 CFR 50.54(a) should be revised to allow a broader scope of unilateral changes to the quality assurance (QA) programs without prior NRC review but that use of 10 CFR 50.59 criteria to make such changes is not appropriate. Instead the staff proposes a direct final rulemaking to modify 10 CFR 50.54(a) to permit licensees to make changes to selected aspects of their QA programs prior to NRC review and approval. Examples of additional changes that could be made by the licensees unilaterally include l adoption of consensus standards newly endorsed by the NRC; use of generic organizational i and position titles; and elimination of descriptive QA program commitments that duplicate those i contained in consensus QA regulatory standards and QA regulatory guides. In addition, the staff plans to develop an attemate process for changes that the licensees could voluntarily implement for further relief. The NEl generally supports the staffs attemative proposal, but expresses concem about the staffs proposed ' monitoring" of the performance of the QA programs. The use of a risk-informed approach to such performance monitoring Eppears to be acceptable to all concemed. We agree with the staff and NEl that the scope of changes in QA programs that can be made . without prior NRC approval should be increased and granting such relief to the licensees should 179
r be given high priority. We also agree that providing this relief through the staffs proposed modification of 10 CFR 50.54(a) is preferable to NEl's original proposal. We are in general agreement with the approach outlined in the draft Commission paper and believe that the staff should proceed with its efforts to revise 10 CFR 50.54(a). Sincerely, R. L. Seale Chairman ! References-
- 1. Letter dated June 8,1995, from Mr. Phillip Bayne, Nuclear Energy Institute, to the Honorable Ivan Selin, U.S. NRC Chairman, regarding NEl Petition for Rulemaking.
- 2. Draft Memorandum (undated) from L. Joseph Callan, Executive Director for Operations, For the Commissioners,
Subject:
Partial Acceptance of Petition for Rulemaking Submitted by the Nuclear Energy Institute (Predecisional). l l l , l l l I l l l j l l i I 180 l 1 ____________________m
%q% UNITED STATES 8 NUCLEAR REGULATORY COMMISSION
$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS %, WASHINGTON, D. C. 20556 November 12,1998 MEMORANDUM TO: William D. Travers Executive FROM: John T. Larkins, Dd Advisory Committee On Reactor Safeguards
SUBJECT:
PROPOSED AMENDMENTS TO 10 CFR 50.47; GRANTING OF PETITIONS FOR RULEMAKING (PRM 5043 AND 5043A) RELATING TO A REEVALUATION OF POLICY ON THE'USE OF POTASSIUM IODIDE (Kl) AFTER A SEVERE ACCIDENT AT A NUCLEAR POWER PLANT During the 457* meeting of the Advisory Committee on Reactor Safeguards, November 4-7,1998, the Committee considered the subject amendments. The Committee decided not to review these amendments and has no objection to issuing them for pubhc comment. Reference-Memorandum dated October 1,1998, from Frank J. Miraglia, Deputy Director, NRR, to John T. Larkins, Executive Director, ACRS,
Subject:
Proposed Rule " Changes to 10CFR 50.47 Relating to the Use of Potassium lodide (KI) for the General Public." cc- J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Colhns, NRR J. Roe, NRR M. Jamgochian, NRR 1 81
# 'o g UNITED STATES
/ o NUCLEAR REGULATORY COMMISSION E I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsuncrow, o.c.aoses November 12,1998 MEMORANDUM TO: William D. Travers Executive s a l FROM: John T. Larkins, Ex Director Advisory Committee on Reactor Safeguards
~
SUBJECT:
PROPOSED FINAL GENERIC LETTER, ' LABORATORY TESTING OF NUCLEAR-GRADE ACTIVATED CHARCOAL" During the 457th meeting of the Advisory Committee on Reactor Safeguards, November 4-7,1998, the Committee considered the subject genenc letter and deaded not to review it. The Committee has no objection to issuing this generic letter. ! Reference Draft memorandum received October 29,1998, from Frank J. Miraglia, Office of Nuclear Reactor Regulation, to Thomas T. Martin, Committee to Review Generic Requirements,
Subject:
4 Request for Review and Final Endorsement of the Proposed Generic Letter Trtled, " Laboratory Testing of Nuclear-Grade Activated Charcoal." cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR T. Martin, AEOD A. Thadani, RES l 183 l
'og UNITED STATES 8 g NUCLEAR REGULATORY COMMISSION $ a ADVISORY COMMITTEE ON REACTOR SAFEGUARDS $ WASHINGTON, D. C. 20656 November 13,1998 i
MEMORANDUM TO: William D. Travers Executive Director FROM: John T. Larkins, e Direct 7 Advisory Committee on Reactor Safeguards
SUBJECT:
SECY-98-253, " APPLICABILITY OF PLANT-SPECIFIC BACKFIT REQUIREMENTS TO PLANTS UNDERGOING DECOMMISSIONING" During the 457* meeting of the Advisory Committee on Reactor Safeguards, November 4-7,1998, the Committee considered SECY-98-253 and decided not to review it. If the Commission directs the staff to modify the backfit rule for plants undergoing decommissioning, we would like the opportunity to review any proposed rule changes. Befpk^ief: SEC'/498-253, Memorandum dated November 4,1998, from William D. Travers, Executive Director for Operations, NRC, for the Commissioners,
Subject:
Applicability of Plant-Specific Backfit Requirements to Plants Undergoing Decommissioning. cc: J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR S. Weiss, NRR P. Ray, NRR W. Burton, NRR l 185
%, UNITED STATES 8 c NUCLEAR REGULATORY COMMISSION
$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g, WASHINGTON, D. C. 20056 November 13,1998 MEMORANDUM TO: Wilham D. Travers Executive Directorfor Operations FROM: John T. Larkins, Execubve DirectorI Advisory Committee On R. actor Saf.guares
SUBJECT:
DRAFT COMMISSION PAPER CONCERNING INITIATION OF RULEMAKING - REVISION OF 10 CFR 55.31(a)(5) AND 55.45(b) , REGARDING THE USE OF SIMULATORS IN OPERATOR LICENSING During the 457* meeting of the Advisory Committee on Reactor Safeguards, November 4-7,1998, the Committee considered the subject Commission paper. The Committee decided not to review this Commission paper and has no objechon to issuing the proposed amendment to 10 CFR 55.31(a)(5) and 55.45(b) for pubhc comment.
Reference:
Draft Commission paper, from William D. Travers, Executive Director for Operations, for the Commissioners,
Subject:
Rulemaking Plan for Changes to 10 CFR Part 55 to Maintain Safety cnd Reduce Unnecessary Regulatory Burden Associated with the Use of Simulation Facilities in Operator Licensing, received November 6,1998. oc- J. Hoyle, SECY J. Blaha, OEDO J. Mitchell, OEDO S. Collins, NRR R. Gallo, NRR J. Colhns, NRR 187
na . I %o UNITED STATES NUCLEAR REGULATORY COMMISSION 3 ADVlsoRY COMMITTEE oN REACTOR SAFEGUARDS I wAsHWGTON, D. C. 20005 November 13,1998 Dr. William D. Travers Executive Directorfor Operations U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Dr. Travers:
SUBJECT:
PROPOSED INSPECTION PROCEDURE 35X.MX, " GRADED QUALITY ASSURANCE" During the 457th rnesting of the Advisory Committee on Reactor Safeguards, November I 4-7,1998, we reviewed the proposed inspection procedure, which provi'es guidance for baseline, programmaticinspection of the adequacy of licensee's implementation ofgraded quality assurance (GQA) programs. During this review, we had the benefit of discussions ' with representatives of the NRC staff. We also had the benefit of the document referenced. The general guidance in the proposed inspection procedure requires that GQA programs have a process for determining the safety significance of systems, structures, and components in a reasonable and consistent manner, including the use of both traditional cngineering and probabilistic evaluations. Further, the procedure provides for a targeted, reactive inspecbon to assess actions taken by the licensees in response to failures of components that are subject to GQA controls. We find the proposed procedure to be adequate for performing an evaluation of the cppropriateness and effectiveness of a licensee's GQA program. i Sincerely,
/ff - ls R. L Seale Chairman
Reference:
Memorandum dated September 29,1998, from Michael R. Johnson, Office of Nuclear Reactor Regulation, to Addressees, transmitting Draft inspection Procedure 35XXX,
" Graded Quality Assurance."
189
# UNITED STATES
-8 o NUCLEAR REGULATORY COMMISSION
$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g wAswiscTom, p. c. 2osss i
November 17,1998 l l The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Dear Chairman Jackson.
SUBJECT:
PROPOSED REVISION TO THE ENFORCEMENT POLICY
' During the 457" meeting of the Advisory Committee on Reactor Safeguards, November 4-7, 1998, we reviewed a draft Commission paper on the proposed revision to the NRC Enforcement Policy. During our review, we had the benefit of discussions with representatives of the NRC staff and the Nuclear Energy Institute (NEl). We also had the benefit of the documents referenced.
The current Enforcement Policy requires inspectors to document any identified violation of NRC regulations. For most Level 1, ll, Ill, and IV violations, the Policy requires that a Notice of Violation be issued. Licensees must respond to such notices, and the inspectors must track and document closure. Currently, a minor violation does not result in a formal Notice of Violation. Licensees are not required to respond to non-cited violations and inspectors close these violations at the time they are identified. The proposed revision to the Enforcement Policy would also allow Level IV violations to be non-cited, but with some identified exceptions. Recommendabons-o We support the proposed revision to the Enforcement Policy and recommend that the Commission approve the proposed revision.
- We recommend that the staff develop explicit guidance with specific examples for implementing the revised Policy. This guidance should cleariy explain the intent and expectations of the revised Policy.
- We recommend that the staff continue discussions with NEl on making other aspects of the Enforcement Policy more risk-informed and more objective.
e We support the staffs plan to monitor and assess the implementation of the revised Policy and would like the opportunity to review the results of the staff's assessment. 191
Discussion The vast majority of violations identified by inspectors are classified as Level IV, most of which are not risk-significant. Since the staff and licensees are expending resources in preparing, responding to, and closing out such violations, resources are not available for more safety-significant activities. The proposed revision to the Enforcement Policy would result in issuing a Notice of Violation for a Level IV violation only under limited, defined circumstances. The revised Enforcement Policy criteria for issuing a Notice of Violation require subjective judgment. Inspectors must continue to rely on their training, kncdge, and experience for determining whether a violation is significant. A review of the inspectors' decisions by regional senior reactor analysts, regional managers, and the Office of Enforcement will be required to maintain consistency and to ensure that the established subjective criteria are followed and the goals of the revised Enforcement Policy are reached. In the draft Commission paper, the staff states that the proposed changes to the Policy will ; contnbute to a more risk-informed process by allowing licensees to resolve Level IV violations j consistent with their safety and risk significance. We note that the revised Policy does not contain any requirement for additional risk assessments and, thus, is not greatly more risk-informed than the current policy. Examples that provide clear direction to the inspectors should be developed The examples should provide the inspectors sufficient guidance for identifying violations such that the inspection is focused on safe plant operations. The staff should clearly define how enforcement actions should change as past practices, associated risk analyses, and administrative requirements are superseded by the proposed Policy. I We agree with the staff that " regulatory significance" should continue to be considered in enforcement achons. The term " regulatory significance," however, should be scrupulously defined to identify those things that are essential for effective implementation of the Enforcement Policy. The staff should integrate its efforts in changing the Enforcement Policy,10 CFR Part 50, the assessment processes, and the inspection procedures to be more risk-informed. These activities are interrelated and changes are not likely to be consistent if made in isolation. Our review of these separate staff activities indicates that little, if any, integration is taking place among the different staff organizations responsible for these activities. The NEl presented its proposed framework for a revised enforcement process that reflects the j move by NRC toward risk-informed and performance-based regulation. One aspect of this proposal is to redefine Level I, ll, and lll violations on the basis of actual and potential consequences that would be derived from risk insights. Another aspect of this proposal is the use of quantitative risk data in the decision regarding escalation or mitigation of civil penalties. The NEl framework is one approach for developing a less subjective risk-informed Enforcement Policy. We believe that the staff should continue discussions with NEl on these aspects, but not pursue additional changes to the Enforcement Policy until after it has gained experience with the revised Policy. 192
3 We support the proposed revision to the Enforcement Policy. The staff plans to assess the implementation of the revised Policy, and we would like to have the opportunity to review the results of the initial assessment of the adequacy of the implementation when they become available. Sincerely, R. L. Seale Chairman
References:
- 1. Draft SECY, Memorandum (undated), from William D. Travers, Executive Director for Operations, NRC, for the Commissioners,
Subject:
Proposed Revision to the Enforcement Policy to Address Severity Level IV Violations at Power Reactors, received October 29,1998 (Predecisional Draft).
- 2. SECY-98-256, Memorandum dated November 3,1998, from William D. Travers, Executive Director for Operations, NRC, for the Commissioners,
Subject:
Proposed Revision to the Enforcement Policy to Address Severity Level IV Violations at Power Reactors. 193
p arg g'o UNITED STATES g / o NUCLEAR REGULATORY COMMISSION 5 E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS "o, WASHINGTON, D. C. 20555
~% , , , . **,f November 19,1998 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
PROPOSED RULE ON USE OF ALTERNATIVE SOURCE TERM AT OPERATING REACTORS During the 457th meeting of the Advisory Committee on Reactor Safeguards, November 4-7,1998, we reviewed the proposed rule on use of the attemative source term at operating reactors and discussed the status of the associated pilot application underway at the Perry Nuclear Power Plant. During this meeting, we had the benefit of discussions with representatives of the NRC staff. We also had the benefit of the documents referenced. The regulations (10 CFR Parts 50 and 100) and associated regulatory guides include a . _ conservative source term (TID-14844) that the staff has considered acceptable for performing design-basis accident (DBA) analyses and for assessing the suitability of the containment design for an intended site. This source term is characterized by the composition and magnitude of the radioactive material, timing of release from the reactor core, and physical / chemical form of radionuclides assumed to enter into the containment under accident conditions. In practice, this source term has also been utilized in other regulatory activities, including assessment of the requirements for equipment qualification and for control room habitability. - New knowledge and experience gained through severe accident research revealed that the TID-14844 source term was unrealistic compared to what would be expected if a reactor actually experienced a core-damage accident of a magnitude commensurate with the DBAs. Consequently, the staff developed a revised source term (in NUREG-1465) with the intention that it be applied to the design and siting of future light-water reactors. The major changes in the revised source term were: an extended timing of introduction of the fission products; a change in the predominant chemical form of fission-product lodine from gaseous 195
2-to particulate; an increase in the quantities of iodine, cesium, and tellurium; and an increase in the number of fission-product groups. The proposed rule would allow licensees to voluntarily apply the revised source term to operating plants. The staff has conducted a number of activities to obtain information for use in deciding whether this proposed rule is appropriate and acceptable, including the following:
- Identified likely plant modifications that would result from applications of the revised source term.
- Sponsored studies at the Grand Gulf, Surry, and Zion nuclear power plants using both the TID-14844 and the revised source term to gain insights on the impacts related to DBA doses.
- Undertook review of pilot plant studies submitted to address a range of revised source term applications to operating plants.
- Performed limited evaluations for the Grand Gulf and Surry plants to determine the risk implications (core damage frequency (CDF), large, early release frequency (LERF), and latent fatalities) of selected plant modifications.
The outcomes of the above activities include:
- The DBA doses are generally smaller with the revised source term (in some cases by a factor of six), implying the potential for relaxation of the fission product control requirements. -
- The effects on the above risk metrics are insignificant.
- Considerable margin exists with respect to the magnitude of the revised source term, compared to the releases expected to accompany a DBA.
As a result of these findings for a very limited sample of plants, the staff has concluded that there is sufficient justification for the proposed rule that would allow plants to voluntarily adopt the revised source term and make appropriate plant modifications. Such modifications would have to be implemented by a license amendment under 10 CFR 50.90 (a new Section 50.67 would be added to provide the implementation requirements). The licensee would be required to repeat applicable portions of the DBA analyses included in its Final Safety Analysis Report to demonstrate compliance with regulatory requirements in the revised total effective dose equivalent (TEDE) form. Because of the regulatory significance of source-term usage, we have had a long-standing interest in the subject. Previously, our endorsements of the staff's source-term related efforts have included: updating and defining a more realistic source term; using TEDE and the " worst" two hours for dose-acceptance criteria; developing guidance on acceptable methods for determining source-term mitigation in containment by natural and engineered 196
3-safety feature processes; allowing application of the revised source term to operating plants (including partial application - particularly the timing of the release of fission products); and developing a better understanding of the risk implications for implementing the revised source term in operating plants. The primary reason for our past support to the development of a more realistic DBA source term was due to a concern that use of an unrealistic source term can result in placing regulatory and design emphasis in the wrong areas. There is also the possibility that risk-significant effects may have been missed or that safety enhancements may have been precluded. An unrealistic source term can result in unnecessarily burdensome regulatory requirements. In formulating the proposed rule, the staff has developed risk information in two areas: The risk implications relative to CDF and LERF. The margins related to the DBA source term magnitude associated with best-estimate DBA releases. The staff's efforts in addressing these two areas of concem have been commendable. The staff has done what could be expected within the constraints of the existing regulatory framework. For the subject rulemaking, it is clear that, to some degree, the likely plant modifications will adversely affect the potential for some quantity of fission product release for plants opting to use the revised source term. For most plants, it is unlikely that the increases in fission product release will be of unacceptable amounts. To a large extent, this is confirmed by the risk-informed values calculated for CDF and LERF. This should be verified, however, for each application. Recommendations in view of the low risk and the possible benefits, we support the proposed rule that would allow licensees to apply the revised source term to operating plants on - voluntary basis. Each application for use of the revised source term should be evaluated with respect to absolute values of CDF, LERF, and the effects of the proposed plant modifications on these risk metrics. Sincerely, R.L.Seale Chairman 197 l
References:
1 Memorandum, dated October 16, 1998, from Jack W. Roe, Office of Nuclear a Reactor Regulation, NRC, to John T. Larkins, ACRS,
Subject:
Transmittal of the Draft Proposed Rule Package - Proposed Amendments to 10 CFR Parts 21,50, and 54; Regarding Use of an Altemative Source Term at Operating Reactors. 2 SECY-98-154, Memorandum dated June 30,1998, from L. Joseph Callan, Executive Director for Operations, NRC, for the Commissioners,
Subject:
Results of the Revised (NUREG-1465) Source Term Rebaselining for Operating Reactors.
- 3. SECY-98-158, Memorandum dated June 30,1998, from L. Joseph Callan, Executive Director for Operations, NRC, for the Commissioners,
Subject:
Rulemaking Plan for implementation of Revised Source Term at Operating Reactors.
- 4. Memorandum dated September 4,1998, from John C. Hoyle, Secretary of the Commission, to L Joseph Callan, Executive Director for Operations, NRC,
Subject:
Staff Requirements - SECY-98-158 - Rulemaking Plan for implementation of Revised Source Term at Operating Reactors.
- 5. Letter dated August 27,1996 from D. Shelton, Centerior Energy, to NRC,
Subject:
License Amendment Request: Revision of Main Steam Line Leakage Requirements , and Elimination of the Main Steam isolation Valve Leakage Control System.
- 6. U. S. Nuclear Regulatory Commission, NUREG-1465, " Accident Source Terms for Light-Water Nuclear Power Plants," February 1995.
- 7. U.S. Nuclear Regulatory Commission, NUREG-1150, Vols.1 & 2, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," December 1990.
9 198
p aw n UNITED STATES 8 o,g NUCLEAR REGULATORY COMMISSION {a r ADV!sORY COMMITTEE ON REACTOR SAFEGUARDS wasWNGTON, D. C. 20655 November 20,1998 Dr. William D. Travers Ex:cutive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 D=r Dr. Travers:
SUBJECT:
SAFETY EVALUATION REPORT RELATED TO WESTINGHOUSE OWNERS GROUP APPLICATION OF RISK-INFORMED METHODS TO INSERVICE INSPECTION OF PIPING, TOPICAL REPORT (WCAP-14572, REVISION 1) During the 457th meeting of the Advisory Committee on Reactor Safeguards, November 4-7,1998, ws met with representatives of the NRC staff and the Westinghouse Owners Group (WOG) to discuss the staff's draft Safety Evaluation Report (SER) on the topical report (WCAP-14572, Revision 1) regarding the WOG application of risk-informed methods to inservice inspection (ISI) of piping and associated Structural Reliability and Risk Assessment (SRRA) model (Supplement 1). Our Subcommittees on Reliability and Probabilistic Risk Assessment and on Regulatory Policies and Practices met on October 29,1998, to discuss these documents and related matters. We also had tha benefit of the documents referenced. The reactor coolant system boundary (RCSB)is one of the primary barriers to fission product release end has been designed to be highly reliable. Piping constitutes a significant portion of the RCSB. Because of its robust de li;;n and the protection afforded by other mitigation systems, piping failures gsn3 rally make relatively small centributions to measures of risk such as core damage frequency (CDF). Assurance of the integrity of primary barriers such as the RCSB is, however, a comerstone of defense-in-depth. Inservice inspection is used to ensure that failure modes such as flow-accelerated corrosion or unanticipated thermal fatigue that were not anticipated in the original design do not unduly compromise the integrity of this barrier. Conclusions and Recommendations
- 1. We concur with the conclusion reached by the staff in the SER that the methodology described in WCAP-14572, Revision 1, can be used to develop risk-intormed ISI programs that will provide an acceptable (and, we believe, superior) altemative to the requirements of paragraph (g) of 10 CFR 50.55(a) and that conform to guidance in Regulatory Guides 1.174 (General Guidance) and 1.178 (ISI).
199
2
- 2. The draft SER identifies changes that the staff believes need to be made in WCAP-14572. /
We recommend that the changes requested by the staff be incorporated into WCAP-14572. We note that WOG has already proposed revisions (Ref. 3) that are intended to address most of the issues in the draft SER. We believe that one of the changes proposed by WOG (Item 19. Ref. 5) should be modified, as discussed later in this letter. We also recommend that the modification regarding model uncertainty (Page 127 of WCAP-14572, Revision 1), proposed in Ref. 5, be omitted.
- 3. Although the codes used to derive probabilities of failure are useful tools, the values obtained are very sensitive to the decisions of the analyst who must identify and select the appropriate input parameters to the code and the likely failure mechanisms. We recommend that the information provided to the expert panel include a discussion of the significance of model uncertainties in code predictions and their potential impact on the classification of pipe segments.
- 4. Because risk-informed ISI can reduce the risk from piping failures, occupational radiation exposure to personnel, and associated inspection costs, we commend the staff and industry for their efforts in resolving differences in a timely manner.
Overall Methodolooy WCAP-14572 documents a methodology that can be used to develop alternatives to the current ASME Code Section XI inspection program for piping. In the Code procedure, the piping is grouped into three broad Classes ranked in order of presumed risk significance. The probability of failure for the piping element is ranked in terms of the design stress levels and the cumulative usage factor. The inspection is focused completely on welds and the fraction of welds, to be-inspected, and depends only on the Class to which the piping belongs. The WCAP-14572 methodology can be used to examine additional failure mechanisms and locations and can provide more informed estimates of risk significance, the relative probability of failure of piping segments, and the number of welds that must be inspected to achieve an acceptable level of reliability. l l In the WCAP analysis, piping segments are classified in terms of high- and low-failure potential ("importance" in the WCAP terminology), and high- and low-safety significance. In accordance with the guidance provided in Regulatory Guide 1.174 and Regulatory Guide 1.178, the quantitative results derived from the plant probabilistic risk assessment (PRA) and other analytical tools, together with input from other engineering analyses, operational experience, and an expert panel, are used in an integrated decisionmaking process to develop the inspection program. The unique features of the WCAP-14572 methodology are its approach to using an existing PRA to quantify risk significance of piping segments, the SRRA model, a probabilistic fracture mechanics tool for computing probabilities of failure, and the statistical model used to determine number of locations that must be inspected in order to meet the proposed performance measure, i.e., a low probability of leakage. 200
3 Upe of Existino PRAs to Determine Safety Sionificance Existing PRAs do not directly incorporate pipe segment failure events. In WCAP-14572, the WOG does not propose modification of the PRA to incorporate these events directly, but instead proposes that the impact on CDF and large, early release frequency (LERF) for a segment can be determined by the use of surrogate events, i.e., initiating events, basic events, or groups of events that are already modeled in the PRA and that have effects representative of those associated with tha failure of the piping segment. Such an approach to the use of a PRA to gain insights on the potential significance of elements not directly included in the PRA could have broader applications beyondISI. The Risk Reduction Worth (RRW) of a piping segment, which measures the reduction in CDF when the segment is assumed never to fail, is used as a quantitative measure of safety significance. Because piping failure probabil.ities are low, if the total CDF for all plant intemal events is used to compute RRW, none of the pressure boundary piping components would be safety-significant, i.e., all RRWs would be equal to 1. To prioritize piping segments, the RRW is instead computed using just the portion of the total CDF that is associated with piping boundary failures. We agree that this approach provides a more meaningful measure of the risk significance of a piping segment. Any application using risk-insights derived from the PRA presumes a sufficient standard for PRA quality. Additional considerations are required when using measures such as RRW. For example, it is often assumed that if something cannot be modeled accurately, it is satisfactory to at least model it conservatively. Although this may be true for measures of overall risk such as CDF and LERF, undue conservatisms in some parts of the analysis can give completely misleading results in the case of measures such as RRW. Both the staff and WOG are aware of such potential difficulties, and until more accurate assessments of the quality of PRAs are available, the expert panel is expected to recognize misjudgments of significance. Determination of Pioino Failure Probabilities The SRRA probabilistic fracture mechanics model used to estimate piping fracture probabilities has been benchmarked against the PRAISE code, developed by NRC. The SRRA modelis intended to be simpler, more user friendly, and more computationally efficient than PRAISE. In a ssries of benchmark calculations, results of SRRA have compared well with those of PRAISE. The SRRA model also includes flow-accelerated corrosion, which is not included in PRAISE. Neither SRRA nor PRAISE is meant to provide detailed mechanistic predictions of degradation phenomena, but used together with insights based on plant operating experience, they provide relative estimates of the susceptibility of the piping segment to failure. The relative ranking will be largely determined by the judgment of the analyst through selection of input parameters to the code. This selection reflects the analyst's knowledge of the phenomenon and operating experience. The SRRA code provides a quantification of this subjective understanding and converts the knowledge that an expert has (the relative aggressiveness of the stressors on a piping segment) into a quantity, the probability of failure, that otherwise would be difficult to dstermine. t 201
4 Effect of Uncertainties Uncertainties include those due to parameter uncertainties and those related to model uncertainty, i.e., the inability to correctly describe all degradation behavior and determine all parameters that affect degradation. The parameter uncertainties, such as the inherent randomness in material properties and flaw distributions, are relatively easy to model, but they are also the least significant source of uncertainty. Although both the staff's SER and the Westinghouse reports focus on parameter uncertainties, the dominant role of model uncertainties is noted. Section 4.4 of Supplement 1 of WCAP-14572 states that model uncertainty " bounds all the other uncertainties, [and] is also the most difficult to predict." The probability of piping failure for systems such as PWR primary coolant piping, where the only damage mechanism is mechanical fatigue due to loads anticipated in the design basis, is very low (leak probabilities are typically <104 and break probabilities are about <104 over the life of the plant). For systems with active degradation mechanisms, the probabilities of failure are much higher (3 to 4 orders of magnitude). Hence, despite the uncertainties associated with these calculated failure probabilities, the classification of the piping segments into those with high-failure potential and low-failure potential should be relatively robust because the analyst and the expert panel need only be able to distinguish those segments in which an active degradation mechanism is present and those in which it is not. The impact of the uncertainties in the failure probabilities on the safety significance classification ' is more difficult to characterize. The WCAP attempts to address model uncertainty by examining - the impact of variations in the pipe failure probabilities on the safety significance classification of the segments. In the SER, the staff has requested that such analyses be performed on a plant- " specific basis to demonstrate that no segments of low-safety significance move into the high-safety significance category when reasonable variations in the pipe failure probabilities are considered. The results of these analyses would be provided to the expert panel. The staff concludes that such analyses would adequately address model uncertainty for the purpose of classifying the segments as either high or low safety significance. We believe that such an - approach is adequate for this application. The WCAP (ltem 19, Ref. 5) should be modified, however, to make clear that the robustness of the classification should be investigated over reasonable ranges of the input parameters describing the degradation modes (flow-accelerated ' corrosion, stress corrosion cracking, vibration fatigue, etc.), since these modes will be more scrutable for review by the expert panel than are the failure probabilities. In its response (Ref. 4) to questions raised at the October 29,1998 ACRS Subcommittee meeting, WOG proposes to address these uncertainties by assuming lognormal distributions with median values equal to the code estimates and the standard deviations estimated using judgment. We believe that there is no technical basis for the assumption that the code results may be used as median values. In fact, model uncertainty means that one does not know how good the code results are. Thus, it does not appear that this approach is helpful. We believe that the issue of model uncertainty is very important and that its importance should be highlighted in both the WCAP report and the staff's SER and that it should be made clear to the expert panel so that the integrated decisionmaking process will be fully informed. What really 202 _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _- - - - - . - - - - - - - - _ _ . - - - - - - - - - - . - - - _ - - - - ---- --. - -- - -~
5 matters is that the final classification of the pipe segments be robust and that the focus of the p:nel's deliberations be the possible impact of model uncertainties on this classification. Determination of the Number of Locations to be Inspected All piping segments, including those classified as having low-failure potential and low-safety significance, will continue to be subject to the system pressure tests and visual inspections currently required by ASME Section XI. The WCAP commits its users to the volumetric inspection of 100 percent of the locations in piping segments of high-safety significance that are susceptible to degradation mechanisms, such as thermal fatigue. Segments with failure modes that have established augmented inspection programs, e.g., flow-accelerated corrosion or stress corrosion cracking, would be inspected in accordance with that program. Other locations in the segments of high-safety significance are selected for examination by a statistical evaluation method that uses the probability of a flaw, the conditional probability of a leak, the frequency of leaks considered ecceptable (target leak rate), and a desired degree of confidence to determine a minimum number of welds to inspect. 4 The proposed target leak frequencies vary with pipe size and range from 1 x 10+ to 1 x 10 / year / weld. These values are slightly more conservative than operating system experience would suggest has been achieved when ASME Section XI criteria have been used. The pipe break frequency, which drives the safety significance classification, is typically at Isast three orders of magnitude lower than the frequency of smallleaks. The proposed statistical cvaluation method has been peer reviewed and determined to be a satisfactory approach for datermining the number of welds that need to be inspected to meet the target leak frequencies at a 95 percent confidence level. Concludina Remarks We concur with the staff's conclusion in the SER that, although the calculation of the change in risk (CDF/LERF) using the WCAP methodology is not precise, it will illustrate whether the result is an increase or decrease in risk. It will provide reasonable assurance that the changes to the ISI program will not result in a total risk increase that would exceed the guidelines in Regulatory Guide 1.174. As we have noted in our recommendations, both the staff and industry have been working diligently to complete the review of the topical report and the Suny pilot project. We believe that implementation of effective risk-informed inservice inspection for piping will be a significant step towards a more efficient regulatory system. Sincerely,
/f1 , A.
R. L. Seale Chairman l 203
6
References:
- 1. Safety Evaluation Report Related to " Westinghouse Owners Group Application of Risk-Informed Methods to Piping inservice inspection"(Topical Report WCAP-14572, Revision -
1), received November 4,1998. (Predecisional)
- 2. Westinghouse Energy Systems, WCAP-14572, Revision 1. " Westinghouse Owners Group Application of Risk-Informed Methods to Piping inservice inspection Topical Report,"
October 1997.
- 3. Westinghouse Energy Systems, WCAP-14572, Revision 1, Supplement 1, " Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed inservice Inspection," October 1997.
- 4. Letter dated November 3,1998, from Lawrence A. Walsh, Westinghouse Owners Group, to Peter C. Wen, U.S. Nuclear Regulatory Commission,
Subject:
Transmittal of Further Proposed Revisions to WOG RI-ISI Program Reports: WCAP-14572, Revision 1 (Non-Proprietary)"WOG Application of Risk-informed Methods to Piping inservice inspection Topical Report" and WCAP-14572, Revision 1, Supplement 1 [Non-Proprietary)
" Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice inspection."
- 5. Letter dated September 30,1998, from Louis F. Liberatori, Jr., Westinghouse Owners Group, to Peter C. Wen, U.S. Nuclear Regulatory Commission,
Subject:
Transmittal of Responses to NRC Opcri he::ns on WOG RI-ISI Program and Reports: WCAP-14572, Revision 1 [Non-Proprietary]"WOG Appi; cation of Risk-Informed Methods to Piping l Inservice inspection Topical Report" and WCAP-14572, Revision 1, Supplement 1 [Non-Proprietary] " Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice inspection."
- 6. Report dated June 12,1998, from R. L. Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Proposed Final Standard Review Plan Section 3.9.8 and Regulatory Guide 1.178 for Risk-Informed Inservice inspection of Piping.
- 7. W. E. Vesely, Reservations on "ASME Risk-Based inservice inspection and Testing: An Outlook to the Future," Risk Analysis, Vol.18, No. 4 (1998), pp. 423-425.
- 8. ASME Research Members on Risk-Based Inservice Inspection (ISI) and Testing (IST) and l Supporting industry Representatives, Response to Reservations on "ASME Risk-Based inservice inspection and Testing: An Outlook to the Future," Risk Analysis, Vol.18, No. 4 (1998), pp. 427-431.
- 9. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998.
- 10. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-informed Decisionmaking Inservice inspection of Piping," issued for trial use September 1998.
204
@ uco UNITED STATES ! o,'n NUCLEAR REGULATORY COMMISSION 5 I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 *%...../
November 23,1998 Dr. William D. Travers Executive Director for Operations U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Dr. Travers:
SUBJECT:
REPRIORITIZATION AND PROPOSED RESOLUTION OF GENERIC SAFETY ISSUE-171, " ENGINEERED SAFETY FEATURES FAILURE FROM LOSS-OF-OFFSITE-POWER SUBSEQUENT TO A LOSS-OF-COOLANT ACCIDENT" During the 457th meeting of the Advisory Committee on Reactor Safeguards, November 4-7, 1998, we completed our review of the reprioritization and proposed resolution of the Generic Safety issue (GSI)-171. During our review, we had the benefit of discussions with representatives of the NRC staff and the documents referenced. DISCUSSION The GSI-171 deals with the ability of the nuclear power plant to respond to a loss-of-coolant eccident (LOCA) followed by a delayed loss-of-offsite power (LOOP). The scope of this GSI was later broadened to include a LOOP followed by a LOCA. The primary concem of this GSI is the possible overloading of the emergency diesel generators (EDGs) resulting from simultaneous starting of all pumps and motors actuated by engineered safety features (ESFs) signals if a LOCA occurred followed by a LOOP prior to resetting the safety injection system. The initial GSI-171 prioritization analysis reported in NUREG-0933 was based on conservative assumptions that resulted in estimated core damage frequency (CDF) contributions as high as 5.5x10 / reactor-yr. Consequently, the staff assigned a HIGH priority ranking to this GSI. Results of this analysis also revealed that nonrecoverable damage to EDGs and emergency core cooling system (ECCS) motors due to surge current created by out-of-phase connection was unlikely. Subsequently, the NRC staff assessed (Ref. 7) the assumptions used in the initial analysis and concluded that the estimated contribution to the CDF decreased by up to two orders of magnitude. To obtain an independent estimate of the range of the CDF contributions for various plant configurations, a further analysis was conducted by Brookhaven National Laboratory (BNL). The BNL probabilistic risk analysis results, which are documented in NUREG/CR-6538, revealed that the CDF contribution varied by up to two orders of magnitude, depending on 205
?
2 postulated plant-specific electrical configurations. The CDF contribution values were found to be lower than those obtained from the initial GSI-171 prioritization analysis; but, for some configurations, the CDF contribution value was sufficiently large to support a HIGH priority ranking for GSI-171. . The staff conducted a survey and also performed an independent evaluation to determine whether any plant configurations were comparable with the high-risk configurations identified in NUREG-6538. The results of a telephone survey of a limited number of plants performed by the Office of Nuclear Reactor Regulation (NRR) indicated adequate electrical load sequencing or load shedding capability at these plants. In addition, the Office of Nuclear Regulatory Research (RES) reviewed the Updated Final Safety Analysis Reports for 20 plants to determine the adequacy of their electricalload shedding and sequencing capability. The results of this review showed that the safety concems of GSI-171 sere adequately addressed in the design of ESF systems at these plants. Based on the above, the staff has concluded that GSI-171 should be placed in a DROP category, and that it should be considered resolved. The difficulties of prioritizing GSI-171 revealed one of the reasons the NRC's process for prioritizing GSis can be time consuming. Two analyses (Refs.1 & 2) and two independent assessments (Refs. 3 & 6) were needed to arrive at a final priority for GSI-171. As stated in our October 16,1998 letter (Ref. 8), there is a need for the staff to ensure the quality and appropriateness of the assumptions used in the analysis supporting the priority ranking of a particular GSI. This effort may require the development of procedures or analytical tools. An important mission of research at NRC should be the development of such tools to aid the staff l in performing these analyses. Although we concur with the decision of the staff to drop GSI-171, we are disappointed with the quality of the assessments performed by the staff. Further, even though the CDF contribution for a number of plants may be sufficiently low, the staff should take steps to ensure that no plant has an electrical configuration that would place the plant in a higher-risk category. In addition, NRR has raised concems that degraded switchyard voltage events at Salem and Palo Verde nuclear plants indicate it is possible that plants have either not implemented under-voltage protection properly or conditions have changed that invalidate original design basis capability. The sensitivity study performed by BNL showed that the dominant contributors to risk from a LOCA/ LOOP accident are overloading of the EDG, lockout of anti-pumping ci cuits, and plant-specific vulnerabilities, such as switchyard under-voltage effects, which may increase the probability of a delayed LOOP and overloading of pumps. However, a degraded voltage condition follow;ng a LOCA is not specifically addressed in the BNL report and sufficient information on operating experience is not available to calculate the conditional probability of a LOOP following a LOCA due to design implementation flaws. The staff should evaluate and resolve this issue through the regulatory process. 206
3 RECOMMENDATIONS e We agree with the staff's conclusion that GSI-171 should be placed into a DROP priority category, and that it should be considered resolved. e Although the contribution to the CDF. associating with load shedding or load sequencing may be sufficiently low for a number of plants, the staff should use an appropriate regulatory process to ensure that no plant has an electrical configuration that would result in an unacceptable CDF.
- NRR's concerns relating to the functional capability of ECCS under degraded-voltage conditions shoula be pursued and addressed through the regulatory process.
- RES should develop appropriate tools for conducting risk informed analysis for efficient prioritization of GSis.
Sincerely, R. L Seale Chairman
References:
- 1. U. S. Nuclear Regulatory Commission, NUREG/CR-6538, " Evaluation of LOCA With Delayed Loop and Loop With Delayed LOCA Accident Scenarios," July 1997.
- 2. U. S. Nuclear Regulatory Commission, NUREG-0933, "A Prioritization of Generic Safety lasues," November 1985.
- 3. Memorandum dated August 18,1998, from Charles E. Rossi, Office for Analysis and Evaluation of Operational Data, NRC, to John W. Craig, Office of Nuclear Regulatory Research, NRC,
Subject:
Request for Review of the Re-Prioritization of GSI-171, .
" Engineered Safety Features Failure From a Loss of Offsite Power Subsequent to a !
Loss-of-Coolant Accident."
- 4. Note dated July 8,1998, from Jos6 A. Calvo, Office of Nuclear Reactor Regulation, NRC, to Brian W. Sheron and Gus C. Lainas, Office of Nuclear Reactor Regulation, NRC,
Subject:
EELB's [ Electrical Engineering Branch, NRR) Determinations Relative to GSI-171. [Predecisional).
- 5. Memorandum dated June 19,1998, from John W. Craig, Office of Nuclear Regulatory Research, NRC, to Brian W. Sheron, Office of Nuclear Reactor Regulation, NRC, et al.,
Subject:
Generic Safety issue-171, "ESF Failure From LOOP Subsequent to LOCA."
- 6. Memorandum dated October 13,1998, from Jose A. Calvo, Office of Nuclear Reactor ,
Regulation, NRC, to Thomas O. Martin, Office of Nuclear Regulatory Research, NRC, l
Subject:
GSI-171, " Engineered Safety Features (ESF) Failure from a Loss-Of-Offsite Power (LOOP) Subsequent to a Loss-Of-Coolant Accident (LOCA)." 207
4
- 7. Memorandum dated October 18,1995, from Mark Cunningham, Office of Nuclear Regulatory Research, NRC, to C. Z. Serpan, Office of Nuclear Regulatory Research, NRC,
Subject:
Evaluation of Assumptions Used in Generic Issue 171 Prioritization.
- 8. Letter dated October 16,1998, from R. L Seale, Advisory Committee on Reactor Safeguards, Chairman, to L Joseph Callan, Executive Director for Operations, NRC,
Subject:
Proposed Priority Rankings of Generic Safety issues: Tenth Group.
"I f
N l 208
/ [ g p UNITED STATES NUCLEAR REGULATORY COMMISSION $ f ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g wasuewarow.o. c. zones December 11,1998 Dr. William D. Travers Executive Dweetor for Opersbons U.S. Nuclear Regulatory Commission Washmgton, D.C. 20555 0001
Dear Dr. Travers:
SUBJECT:
OPTIONS FOR INCORPORATING RISK INSIGHTS INTO THE 10 CFR 50.59 PROCESS During the 458" meeting of the Advisory Committee on Reactor Safeguards, December 3-5, 1998, we met with representatives of the NRC staff to discuss opbons for incerwe;ng risk insights into the 10 CFR 50.59 (Changes, Tests and Experiments) process. Our Subcommittees on Reliability and Probabilistic Risk Assessment (RPRA), Plant Operations, and Regulatory Policios and Pracbees met with the staff on August 26 and November 19,1998, to discuss this matter. We had the benefit of the documents referenced. We appreciate the opportunity to review the staffs work during the early stages of development. We recognize that the staffs approach is still evolving, therefore, we offer several observabons and recommendsbons without commenting on the detalis at this time. Observations and Recommendabons
- 1. The objective of this work is to develop options for making 10 CFR 50.59 risk informed A key part of this effort must be to determine which attributes of 10 CFR 50.59 are better served by the use of risk information and which are better left alone. The description of, and the problems associated with, the existing process should be used at the outset to identify where risk information can enhance the process.
- 2. Any changes to 10 CFR 50.59 must both preserve and improve the desirable features of l the current process. The staffs report should start off by articulating clear measures for j improvement and constraints imposed by other regulations or requirements. The draft report containa several useful evaluation factors. However, the term " enhanced safety .
decisions" is not a useful criterion unless it is made clear in the context of the safety ) objectives of the NRC. A difficulty with 10 CFR 50.59 is the lack of a clear basis for deciding when changes are l allowable without prior NRC approval. We recommend that the staff evaluate how risk ; information may be used to address this problem and that it be documented in the ! beginning of the report to guide the development of viable options, and consideration of I new attematives. 209
2 The constraints on allowable changes need to be expressed clearty in one place. Currently, references to inadequately defined terms, such as " safety status,*. " adequate protechon," ' defense-in-depth," " safety margins," and "operatonal safety," are scattered throughout the report. If they are to be used as constraints on the options, they need to be gathered, defined, and hated so that they can be used in a more formal way to evaluate all opbons on a common basis.
- 3. The staff has considered several opbons and made preliminary evaluatens. Although some features of these options appear desirable, none emerges as a clear canddate for implementaten. We recommend that the staff reconsider the selechon of options, involvng creative combinations of the best aspects of those originally considered as well as other, bolder opbons, such as allowing changes that do not affect the technical specdications, or affect the technical specifications but satisfy criteria similar to the extension of the Regulatory Guide 1.174 criteria, as suggested in the attachment to our j I
July 16,1998 report.
- 4. We are concemed that the staff does not have sufficient time to property evaluate the options. We believe that more time will be needed after this preliminary effort to evaluate the candidate options in detail. For example, we would like to see a set of test cases that verify and validate the mechanics of the competing approaches. Due to schedular constraints, the staff developed only one test case regarding the reclassificaten of the South Texas Project essential cooling water screen wash booster pump.
- 5. The staff should consider the issue of combining changes. In the report, the staff cited examples of individual changes that resulted in unreviewed safety questions, yet, when combined with corrective or compensatory actions, the overall change would meet risk-informed criteria. We recommend that corrective or compensatory actions be given more prominence in allowing collective plant changes using risk insights.
We look forward to discussing this matter with the staff during future meetings. Sincerely, e . #V - R.L.Seale Chairman References-
- 1. Draft paper received November 20,1998, from the Office of Nuclear Regulatory Research to the Office of Nuclear Reactor Regulation,
Subject:
Options for incorporating Risk insights into 10 CFR 50.59 Process (Predecisional).
- 2. Report dated July 16,1998, from R.L. Seale, Chairman, ACRS, to Shiriey Ann Jackson, Chairman, NRC,
Subject:
Proposed Revisions to 10 CFR 50.59 (Changes, Tests and Experiments). 210
/ UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION l f,,
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
$4 WASHINGTON, D. C. 20065 l ess**
December 14,1998 I
\
l Dr. Wilham D. Travers Executive Director for Opershons l U.S. Nuclear Regulatory Commission Washington, D.C. 20555 0001
Dear Dr. Travers:
SUBJECT:
PROPOSED COMMISSION PAPER CONCERNING OPTIONS FOR RISK-INFORMED REVISIONS TO 10 CFR PART 50 " DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES" l During the 458* meeting of the Advisory Committee on Reactor Safeguards, December 3-5, 1998, we reviewed the proposed staff opbons for making risk-informed revisions to 10 CFR Part
- 50. Our Subcommittees on Reliability and Probabilist;c Risk Assessment, Plant Operations, and Regulatory Pohcies and Practices have held four meetings since August 1998 to discuss NRC staff and industry activities associated with making Part 50 risk informed. During our reviews, we had the benefit of discussions with representabves of the NRC staff, the Nuclear Energy l Institute, and other interested parties. We also had the benefit of the documents referenced.
We recognize the importance of revising Part 50 and of resoMng the associated policy issues. Because of the staff's tight schedule for completing these tasks, we are holding frequent meetings and are making use of early draft documents. Since the staff's approach is still evoMng, we offer our preliminary views on this matter without commenting on the details of the draft document. In our recent meeting, the staff presented a number of options and policy issues for making risk-informed revisions to Part 50. The policy issues identified by the staff were: a voluntary versus mandatory conformance with the revised Part 50, I exemptions for pilot plants from current Part 50 requirements, and l I *- modification of the scope of the Maintenance Rule. Our view is that there is really no choice but to make the revised Part 50 voluntary because a { backfit analysis will be virtually impossible to do. Making the revised Part 50 voluntary, in a sense, connotes " proliferation" of regulations. As noted in our September 30,1998 report, there will have to be two sets of regulations - the current rule and the revised rule. Although this < should be recognized, it should not be made an issue. 211
2 Pilot plant studies should be helpful in identifying the range of plant changes that might result, along with the impacts on risk, resources, and on other regulatory metrics. Exemptions from the current Part 50 requirements for pilot plants will be an expeditious way to better understand the implications and consequences of the revised rule. We are not yet prepared to express our view on modifying the scope of the Maintenance Rule until we gain a better understanding of the relationship to other regulations, especially the License Renewal Rule. We support the staff's recommended option of a two-phase approach, which first addresses the definition and scope of " safety related" and then proceeds to a second phase of revising specific sechons of Part 50. We believe that the redefinition of " safety related" is fundamental and necessary, and is a strategy that will produce the most positive effect in the short term. Before the staff proceeds with the second phase, we recommend that the staff more completely define certain terms - particularly ' safety" and " adequate protection " We also recommend that the staff better clarify the principles and policies for defense-in-depth in order to identify necessary and sufficient limits. The staff should determine which elements of Part 50 constitute defense-in-depth and which of these should be retained. We believe that the lack of a clear articulation of the functional ob loctives of a risk-informed regula6ry system is contributing to the difficulty in identifying and evaluating options for making the existing system risk informed. Satisfying the Quantitative Health Objectives, which can be measured by subsidiary goals such as the core damage frequency (CDF) or substitute goals such as the large early release frequency (LERF), is not the only concem of the Agency. Additional objectives, such as the preservation of defense-in-depth, safety margins, and
)
operational safety are often cited by the staff. Articulation of the regulatory objectives should be a restatement of the NRC's mission related to what it may consider to be the complete definition of safety, as contrasted to risk. For example, during its work on revising the inspection and assessment programs, the staff has identified a number of"comerstones" that could be adopted as the objectives of the Agency. If all of the Agency's regulatory objectives were clearly stated, the requirements for making Part 50 risk j informed could flow directly from these objectives. ! Sincerely, i R.L.Seale Chairman References-
- 1. Draft Commission paper for the Commissioners, from William D. Travers, Executive Director for Operations, NRC, received December 1,1998,
Subject:
Options for Risk-212
3 Informed Revisions to 10 CFR Part 50 - Domestic Licensing of Production and Utilization Facilities (Predecisional).
- 2. Report dated September 30,1998, from R.L Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Impact of Probabilistic Risk Assessment Results and insights on the Regulatory System. f I l 213
# %, UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION $ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g wassmorou. o. c.aoses December 14,1998 MEMORANDUM TO: William D. Travers Executive Director tio s FROM:
V I-+ w John T. Larkins, Ex e Director Advisory Committee on Reactor Safeguards
SUBJECT:
RULEMAKING PLAN - PROTECTION AGAINST DISCRETE RADIOACTIVE PARTICLE (DRP) EXPOSURES (10 CFR PART 20) During the 458* meeting of the Advisory Committee on Reactor Safeguards, December 3-5,1998, the Committee considered the subject rulemaking plan and decided not to review it. The Committee has no objection to issuing this rulemaking plan.
Reference:
Memorandum dated November 13,1998, from Jack W. Roe, Acting Director, Division of Reactor Program Management, Office of Nuclear Reactor Regulation, to John T. Larkins, Executive Director, Advisory Committee on Reactor Safeguards,
Subject:
Rulemaking Plan - DRP Dose Constraints cc: A. Vietti-Cook, SECY J. Blaha, OEDO J. Mitchell, OEDO A.Thadani, RES A. Roecklein, NRR S. Collins, NRR 1 215 L
UNITED STATES .8 %,, NUCLEAR REGULATORY COMMISSION
'd I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsmucTom. o. c. moses December 18,1998 Dr. William D. Travers Executive Drectorfor Operabons U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Dr. Travers:
SUBJECT:
PROPOSED IMPROVEMENTS TO THE NRC INSPECTION AND ASSESSMENT PROGRAMS -INTERIM REPORT During the 458* meeting of the Advisory Committee on Reactor Safeguards, December 3-5, 1998, we reviewed the proposed changes to the NRC inspection and Assessment Programs, including initiatives related to development of performance indicators and a risk-based inspechon program. Our Subcommittees on Reliability and Probabilistic Risk Assessment, Plant Operabons, and Regulatory Policies and Prachces also reviewed this matter on November 20, 1998. During these reviews, we had the benefit of discussions with representatives of the NRC staff and the Nuclear Energy institute (NEI). During our 456* meeting, September 30 - October 2,1998, the NRC staff and NEl discussed the results of the Performance Assessment We&#,ep held on September 28 - October 1,1998. We also had the benefit of the documents referenced. In our September 10,1997 report, we recommended that the staff utilize a top-down structure to develop the assesstnent process. The staff has adopted such an approach, beginning with a statement of NRC's overall safety mission, progressing to a set of Strategic Performance Areas, tnd resulting in the development of seven Comerstones of Safety. The staff established three Task Groups to propose improvements to the inspection and Assessment Programs by making them more objective and risk informed: A Framework Task Group to develop details of the framework for a more objechve, risk-informed, and performance-based approach to assess licensee performance and related bases forinspection activities. A Risk-informed Rebaselining Inspection Program Task Group to develop a risk-informed baseline inspection program to assess nuclear power plant licensee performance. An Assessment Task Group to develop a process for use by the staff to make objective conclusions on licensee performance, take actions based on these conclusions in a predictable manner, and effectively communicate these results to the licensees and the public. 217 l
2 The staff has made significant progress since our March 13,1998 repod to the Commission. The staff is finalizing the Task Groups' reports associated with the overall framework, inspection, and assessment-including the performance indica' ors and thresholds for regulatory action, and the Transition Plan. The staff is developing a set of performance indicators for evaluating normal power operations that are related to risk by accident-sequence logic, are supported by objechve data that are readily available and scrutable, and are directly measurable or calculable. The relationship of the performance indicators to the risk-informed Inspection Program has been well<$eveloped. Development of performance indicators for the areas of shutdown operations has been proposed We understand that this work will take several months to complete. The approach of the Risk-Informed Rebaselining Inspechon Program Task Group appears to be sound. Although the baseline inspection program document is not yet complete, its ; organization, format, and content give indication that the staff will develop a comprehensive inspection program that is successfully linked to the Comerstones of Safety, l A Transition Plan for implementing the revised Inspection and Assessment Programs has been { developed. This Plan contains such elements as milestones, training requirements, and i communication plans that are essentiai to a smooth transition. Each of the Regional Offices will have knowledgeable personnel for implementing the Plan. The Plan also provides for several pilot programs that are essential for successful implementation of this complex project important elements that enhanced the progress made in developing the new inspection and Assessment Programs include:
- The use of a top-down structure to develop the assessment process.
e Improved coordination among the various NRC Offices, and the Regional Offices.
- The involvement of stakeholders in a well-organized public workshop and the continuing interactions with the nuclear industry.
Although substantial progress has been made, much work remains to be done particularly with regard to the integration of the assessment process. We plan to continue our review of this matter during the February 1999 ACRS meeting. Sincerely,
. . A R. L Seale Chairman 218
__ __ _ A
3 References-
- 1. U. S. Nuclear Regulatory Commission, Draft report, "NRC Power Reactor Baseline inspechon Program," received December 1,1998 (Predecisional), including draft NRC report sechons:
- Barrier integrity Comerstone," dated November 18,1998.
" Mitigating Systems Comerstone," dated November 23,1998. I
- 2. U. S. Nuclear Regulatory Commission, draft report, Revision 1.1, prepared by Los Alamos National Laboratory, "Results of the NRC's Assessment Team Task Group Working Sessions, November 1998," updated November 11,1998 (Predecisional).
- 3. Memorandum dated November 4,1998, from Ashok C. Thadani, Office of Nuclear Regulatory Research, NRC, to Samuel J. Collins, Office of Nuclear Reactor Regulation, NRC,
Subject:
Draft Recommendations on Development of a Risk-Informed Baseline Inspection Based on Review of Individual Plant Examinations (IPEs) and Probabilistic Risk Assessment (PRA) Insights" (Predecisional).
- 4. U. S. Nuclear Regulatory Commission, Letter report JCN W6234, Draft #2, prepared by Brookhaven National Laboratory, " Development of a Risk-Informed Baseline Inspechon Program," dated October 28,1998 (Predecisional).
- 5. U. S. Nuclear Regulatory Commission, report prepared by Princeton Resources Associates, " Plant Specific CDF Information to be Considered for Risk-informed Inspechons, Surry 1," dated October 23,1998 (Predecisional).
- 6. Memorandum dated October 13,1998, from Thomas T. Martin, Office for Analysis and Evaluation of Operational Data, NRC, to Ashok C. Thadani, Office of Nuclear Regulatory Research and Samuel J. Collins, Office of Nuclear Reactor Regulation, NRC,
Subject:
Risk-Based Performance Indicator Development Program Plan (Predecisional).
- 7. Nuclear Energy institute, Draft report dated July 9,1998, "A New Regulatory Oversight Process - Toward Risk-informed, Performance-Based Assessment, inspection and Enforcement"(Predecisional).
- 8. Report dated September 10,1997, from R. L Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Staff Action Plan to improve the Senior Management Meeting Process.
- 9. Report dated March 13,1998, from R. L. Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,
Subject:
Proposed improvements to the Senior Management Meeting Process. 219
- NRC FORM 335 u.S. NUCLEAR REGULATORY COMMisSloN 1. REPORT NUMBER l(2-69) (Assigned by NRC, Add Vol Supp., Rev, lNRCM 11o2. and Addendum Nuncers, W any.)
22ai. 3202 ECLIOGRAPHIC DATA SHEET tsee estnesons on en nnese)
- 2. TITLE AND SUBTITLE NUREG-1125, Volume 20 A Compilation of Reports of the Advisory Committee 3. DATE REPORT PUBUSHED on Reactor Safeguards: 1998 Annual MONTH YEAR l
April 1999
- 4. FIN OR GRANT NUMBER A AUTHOR (S) 6. TYPE OF REPORT l
Compilation !
- 7. PERIOD COVERED pnclusve Dates)
Jan. thru Dec.1998 8 PE JORMING ORGANIZATION - NAME AND ADDRESS (t/ NRC. proude Dawon, ornce or Repon, u S. Nuc8 ear Regulatory commessen. and madng addness. # contractor prowde name and mahng aQdress) Advisory Committee on Reactor Safeguards { U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
- 9. SPONSORING ORGANIZATION - NAME AND ADDRESS of NRC, type "Same as above*, # contractor. provele NRC Dwson. Otsee or Repon, u S. Nui:ssar Regusatory commesson, and madng acMress)
Same as above l
- 10. SUPPLEMENTARY NOTES
- 11. ABSTRACT (200wordsordess)
Thb compilation contains 59 ACRS reports submitted to the U. S. Nuclear Regulatory Commission (NRC), or to the NRC Executive Director for Operations, during calendar year 1998. It also includes a report to the Congress on the NRC Safety Research Program. In addition, a report to the Commission on the NRC Safety Research Program, NUREG-1625, Volume 1, is included by reference only. All reports have been made available to the public through the NRC Public Document Room, the U. S. Library of Congress, and the Intemet at http://www.nrc. gov /ACRSACNW. The reports are organized in chronological order.
- 12. KEY WORDS/DESCRIPTORS (ust words or phrases that me assst researchers m Jocatmo the seexrt 1 13 AVAILA61UTY SlAILMLNT Unlimited 14 SECURITY CLAS$1FICATioN Nucle:r Reactors Safely Engineering crus eages Nuclear Reactor Safety Safety Research Unclassified Reactor Operations , g, Unclassified
- 15. NUMBER OF PAGES
- 16. PRICE NRC FORM 33S (249) TNs form was electromcally produced by Eine Federal Forms, k
l Printed
- on recycled paper Federal Recycling Program
. - $l
, * 'l h .l, .: ,,< ' ' . ' ' . . * ,- I' < ),h. ?!'. l. 't
. ' , ' j. 5, y \ ,* :. l- .f,4l?.,,*:' . .f;, f'.h,5 ,j *k - .*' .: j 5 a.' ~..I',,l" ';^
l " Y . .', I. ,* h, ; . ,l 7, .ft. :;; .,,
- lt .'y.,.,'.,'.v.j ,, .y,
,,, , .,y,, pg ..a...y,.i.,...'.,..,,"..*., . < , ,* ..-}.<, .., , , , , ,,. . , . .. . r, . .v ., j j.,j
. $J g ec.,....,,.A,.., .s,< .7, . ,..3 , .. .: ~ . .
g
' ;;< . jg..;7 .. . ;
, ., i. ,,,p.,y,.
j
;, . ;, ,,, ,s.... g..
. . . .:;. p..
</ .[,
~ 'l. '.,I.[ I,h .
.Jts , t
; r ~ .!Q ,, "-[. , ; ,[ b ',' l[ $ ,
,] , . I,'p l3 ,e7.,N; ,;
%,e, ,h'..
1
.,,[,,,
?.g...,.....h.,.'
, ' f;). t
". ,f ) ,;
if.',.? ..[' (! - . . . l~ .' n. .
;-l Li Y, E : -G,)f,[a .y
., n,,p v. . , . , a.,:v.:.".. \..!.;<n
.n. y.u..;;. . ? v',: t ' . .. f. . ,.3. .;^ , . ;. 6,, ~j . 4 ::, " : ., '.; u :i.
.'4 ;-
1 .
... . . . .r-
, . i
. .f. ~..; ..v .:s-, . ..
- n. :. ..:+; .
o , .O. .;. 4, r., .i y, , :: ; p. ,Q .,.@<, ,'.::. , , ,. .r:, n.. . ,.. , u.::
..,. - - .- . .. 4 . . o. . - -
..m , c % ..,
, .\.,,*: .f'i ,I75l i:\l.:.'.v.,.}M:k;...':.'. ll ....'",,~. e'". ;;f' e,h. -['['Y.
J. (( 5$n.n.(.$'ll[,.1~ a..
-[,,.O,T!Y.M'N,1.!M- }il2.)i-N/[.l:kU.S'[h:: . . . .
. . v. .' ....o.,
.7..
..c., . . .. . ,; .., . -.w .~. ..o. . . r . . . . ; . . . .:. u,....,: v.. ,,*v. iy: . .; ..
s . . .u.
,.,,.,e.. .. , .
, s . <. .
~,.'.,...c.
... e . . ....
,. ..+, .;',. . . .;
. .: . .. . .M .' ....
. ,. . p,) , . , , ' , ,' .. ,.: s
.'r .,,',9 .
..4, *
- p. . , , . . . , .. . , ' . , .
v.' ..'.,.',i':'.1 ,, :s
.,...,...'.'?,,L:n ... .'. .* : . . 's: . l';s-a
.G',... J .: .
; .- ' ... . .. 9 . . . . . . .,, . . , ... ;-
1 ; ; 's , .4 , ;.~; ,.z, .,.,.'~.' . .n ;'.; .. o,. : . .. . . ; s :; , ;;.;; .'f.. a 9. . v .:, , ;. . .:; ..y.z. y . _.;,.p. g.,:, m,t. ::..;. . . . :: . . . .; ,;?,
..- .,.- , r+..:'<.. [c
. c ., . ,. . n .. . -n. , n.c._..
- q. ;...'.e;.,.l.:
. . ~ n . . . . - ,..,_ . . . :~y :. - .. . . ..c., ...
*.,.,.t .uL ,. :.. 4. . ..
- e .. .
- . :p ?9...., 4. y. . . 2 . . . .: .,.p...'. : :L,, . . . . ? . .' : y, ~9.:> < : -' . . . . . . . " . ,. ';:; ".::. .s
, -, ' s :. ..*: .,': . .: .6 :: .u, n .m,....,<,,'i._:>,;.~.. ':n,
- s..:< : '.: ,s.+ ., ; q y :;. .a. .?. t
.. , . .s . . : ..
,.g",..'..,. s. '. *......,,':t
. . . , .* . c. -: . ' , . .* .*<' ., .
.t'...,',.s.s ..',,..t..,,../
o..,*.s,.c
.< 1.
,."<.,,lg 4. , .1.'
, . 3- '
,, . ...i , J, c.
- .p .., . ..>. r .q ~- ..m,......#,j.,, , , . , , , , . .;
8
.t' r';M,.,,... z, n' , #- ' , . ' .. ';'..'),.
; y ,. . .\ G , ;,.,,,'pf. J ' . .,. . , ' ,- ;,+ . " ., s s . .
- It
- '..' ' ' , , * -. . ' {%.' ', *
'.i. .:y '; r ;-lr'.,: . ~'
&; : h. % ; .h. f,,;; };1;.n . .;,% , ' . '.'.....,...,'w' ,' G,,;. l,ll.,r ., .'.'-'.or . , : . %.3;. . , ,
r
. . . ,," ,o.
.,n-[~-
. A .' =.g. ,\. ' ';":n, (,
. , .' ,. ;l-r ,g" .s, *. > ..,..,2.- <,'~.s**..f,,.,,..,*.,,j'.:,%.'...*:.
.., s c
't
- y. . .. ,,.o:;..., ,**. -.-
..,.,,e, t
...j,.s<..,s.,,.
,e.. .,.. .. ..
%. . -?,'...[.,,.'
- i. ..
>.; 6...,,v.. ...~...,
.,._.v ;.,4. ,. ,... '.;v. . ,., , .-. .n
,...e~ . . . .. . . ..a. ~ :y.'>
- f. .. . ,
-..a'. ,. .. .... <
1.
)v. . 'f...,.
. ,. ,' ,, b,m.-
. .r. .... c.. .
a.,. .
. :r. . ~ ',; W * .*. .,.h,.,$+ * - . Y.f . ..* ? , '
S ll!g 4' '. . f .s '. 'Y1.. l
,.. r ., .*. ': ' .i, ? ! :'
. $,,t,,ll . ,a' Y;;,[ .: . , , .,{ . b-l.Y ', rl .'. N.' ' k. .' . . ' A. ;,,, '. s' *) -l, ' . [ , . , ,'.i,. ...,....,'.,.,n.?;, ?.
.: a t.y..y.,,'._.
. .., + .
..,...,,..'.;;'...g:-' ,i ; ....[i.p..., .* -
3
,._,.t'..,... '. ,,.i'.- q:
. , .'.y V:t..f,. , . - ' , .. .f . .
t.' j , * - y....><..',, l ...,. , . , . , .r._.,.'.n.. , ' . ' . . . **:., ; , .. .n' ,. : l v' *, ?,
,.'.*V'Q.*'t.'.,q.f . ..,:. .',ej<. .s, . a . '?.,Q..
, . ,t w
.,....s... . . .. . ' , . . . . . . . s .
.,.)...s',....;.
-l*., s.,~.
% f , ,'\ . ,' a,' '%.:' . . ' ' '. . ' '; 9 * . . . - *, ' . . - . , .. -','[ , , '
. . .r .;;., * '. ,. ~ .t.- .- .', l f. , .? .l . .. ; - ; ~; g. :
, n'.,....;- . -'- .; .(
n
' '.(*. L ? '.
...l . :. ,,. .. .] ;:,,R,'. 3. d ,L.[.' f _; :...,. .: ? - ',,,,..?e-y '.L . ,, .. ., :;;a y ;. . * ., , ; ..'A
...:._._q'-l...;1e'..
,. ; .;';, ~:
's ,',l : .
- h'R
' i;-]?
,W'- , .; k , _ ) .;p. "J'l.'., .y ,',. * ,' ;$. h" 2,,,..' * -
f, . ':.] .y"t ..g,?. . .
' '. s. ,.,' ,' ;(; , $ e e .[, 7,: - , '# ,', $ c , ' , , ' ,'/[ ' . ':: c4
$<; ,{. ' i. ., ' .;; g, .
',. ' . 7. j l . ,f _ ., 3.s ,s y O. .,'.-.
,.- b
.', d: \..:.' + r 4 t,4, Q -'f. , v , f .;', o, 8% : ,;4,**-
'., ..y';*~',,';*', ;
~ , ;. ,
' a 5
l,'J', W, .
.v'! ' ' ' 'a, .'. . h',k p..,'s',..'.!
~
p ' # ' " E.'li % ,' d
..'.,'.'.I -
- e. .[, , '
;*;_ ' 6'.h t#e iC
, (,
','}' * '#'
,,h..,
e., 4. : - ,,'i,
' l . ?. ao r.,,
,9,:*....,,f.,,,'.'.,,gd'g".-,\'..'s.:'b ..
.s .. i *: ' .n.s . . ,p ':. . g n
>.e s .i * ' l - ', r
. P: ',. - r, j} y ,g( ; y , ,. 3.l) g ,,i,f .
- l
' A : .,,. k, f;- ' ff: '
., ,, .. ,' [. .,. .. .6 * ]g
.Js./,,,...-+
. y.. .,,.,j,,...,;e
.':,- * [ ,. l..'- .',,,, ,., ..j*'.' . . .,., J[8 *, . ,
...y- f a + < , .-
. y,
, . ' ' . " , ,,l, l (,,f , . ' / , s';
': -u
,,,f,g,,
.. i-
; N
- g. . . .' . .'. . . ~"u.f. ,, , . . . 4.
'.y,,
. .., : .i ,-
. . , . . ' . . ,, , . , ;i p
f t,s .
,.' " .' y .*:; . ,.i, ^, . . ~. n ,, . , . .
' r/ s,'v.,2 ,.e.,
- .
-: l ;. . ~ . . {s . . . , , . . - . ., c . . .~ .': . . ", . . 4'.z: . : -n -.f..- - , c w . . sas . v. ) . ,,'y. I'; * . E f +f <,[. ;, ) , ~' '1 'p* *, ' '.';'.. .!7. . ,#. , ,. 7. [ ,el 'l , . is. o .I. . f. .}3,, .; . (7 ., - . Ii':,..,r 't , . ' , ... ,[ ' y j - ,g. ;, t * . " . . , , , ' . . , . ' ' , . 't t,.; ' . -, ., , . .' H. , *1 ", ; , ,,1 n . -{,h, i.. ~ . r' , 2 :. . . " :: .
- s.. g _. .,,',
, . - - . , ...:, .r . - <;n .;['; .:,y 3 q,3 :. . L ;. . . ';;c. f. . .m . r,;, - , ;.a ,p.,3.y~ . y..., , ',}.' . ( ;. _ ;,n_ , p,4,?. .-, . . ~ :':.; . . , . ; . .> p . .'; c .,. , :;.d ,.; p, :,:. ; . ','.y. s, : ... . , ,,: . _ ,.; - .,.y- ' yL;,.. , , ,,.:., . . . . _ .p :s~ -c ';;;.};, .:, .. . .e . ..e:. ._..,,., .*3. ... ..,,4, . n. ;,. y. :, ,f. ': _;.., .- ^g. '((, ,., .';.Q \ N. ( l : ,0.. .f.,. i . f., '?, .,r,:[ ,: .[::l:a :,? ;I',i'l _",.W;:,. . .. [ ,..; . ;. : ,, , '.., ?. e , . .,. . L lt .d .i ;_.:. _ . , 4; , . %.._,.l;3,,,' s - . . . g ,l C;.Q,. 5 . .' f' ' f, , l.]~..~,'Y , . . . . . ., . l :',;' . ,,:;:A'. . Ee.g. .d.l l t ,t. . - *; .a.+,<,...,.,<,..r '*,,.. A.,...s.. $.i. . . - Y,l. 'o,. lll .a. h ,.-' '.' . e'.; .] '. "), ) .A.T. ; , ' .. . .. n , . , , . . .'. _ - % .4s,, .r.. .;- . * . " ...,*y......-. . -,. > , . 4 s... . @ ... . . , ,.- .... ~ - .. .. . i,, ' , - . , , . ~ , :, '. A . ~ ., . ,_ . . . ,.. o . . . . .. . , . ...g y ,. r. ,\ .
- y
..,.,.,:<,M. s < .r...
- s. .
e...'..., 4 , .~.. <*.',.,,.s, , . . .....4...
- ....a. _..y* '.. .. .,f,;;..,,..o,........ ,..,v..- . ..,y....
.-. .;, ,.. , ,.. .', .J,,1._.,,*;'l,...' .' a,' ..:. .,~ . .. v. e- t,v.. . : , y_ .,.t s . ,,,e .. * , ;.,... . . %. 9. . ,,.~ . , . , . ,S.*, ., .. n ...s, .,- *, . -..,;
- , .,,,..,(,',"
; . . 7, .; 1- ,' ,, ; j':l ,'.4 *'y.4, l.* ,' . j [ 1, { . .' e, :<> . .. ;. ' -] b* , , f ",',,'.,=^>.I,.',-J<'..,:,I.s. .. . ,- , ,5 .. s . 1p. , : . . ,.. .s ]*..:, .. , . , .- .,,s. . , . ,. . . , . . %.e,s.... . .:,. .,. , , :j .. . . . , . , . . . . , ,s .. , L. . . ,. . . u,.,.v ; - ...... ., .,t ,, : ,".,.;,..,.....,'~ . ,. .t ,:: '~ ...3., - -t . .. . . . . , . ;. 4+ q . :" %. .'. ,s ,, . , ,. .;. ,., -, =. , ,. . - :: . . >;. :'A .< *:
- , ',,,.*,.~3.., ' ; c..g'," :..;.'.. . ..' . .
;. . . y . / . : 'a'....,.1.,',...'...g.,.g....1.5:f,. . f. :9 a.... .. ., .: .: *,<. 6 , - : .i f; r ; . f,: . . . . . . . .- -
- ,. . s,
. w; 3. 3 . ;m,.., . .y ,- n . . :;e , r ..- y,..,.n u .*e. . '- s ; - ....a,, .. ~.s , .- .1 .' ' . ... . . . , . - ',} . . .4.: , ' . . ' . ' ' E h/.' . . , ' ' (, y' . , ,- ,. ' ' , f'l . ,, . ,, :. ., '.,[,+'", - ,,'T,' I*,t.....',. j[- fff' , , , .. . N'. . ,, f.y ~N ,1 *.( ,. f ,, I'. * -o . - .., '9.z., N. .::.. > , .V~ .; -,'.?. , . .. ., ' . n .,, , ;\ . . " .: . . .: - . . ^ .,: . . .: ':. .)"a , .,. . ,s.,.*,5.- .: . . , , . . , , ' , . . . M' ', ,.* .., . p, , "y, . ._?.,. . . ;: ,. . : .'.._;,:.,.. '.'( , !,'.:.}' ! : ' q*:......
- r ,. .I ..a ' ,,% . ~;
'.,. .: . ,4 , . .- .;, ,' - , : , ; . e.. =; .,' , r.* ,.: 3f,. ,. . . '; L . n : c i, , % - '.tt . . .q. . '. ;.;..*...,it+,:,'. - . < .., w . >* . , . , ;,, . .. . :- R,): ,; <.z -. ss^ '-, e~, ,.:- ~ ' . . )".; -; .,..~ "f: ' ' , 4. , ', * , , , . . ' . - j . . < - ) .e ,s : ' . - ~ , ..',,..'.:l,q, y,,,.i',,,.1."4.,,),. ,'8 r.Y ;; R ,_ , .% l .( . .; . .. ;, - *.c : g: , " g ;.7 . , 1j. . ; .~. ,2 ,) ,*, .' , ~ . g * . -- (* , . .. S , ,'-: .: _3. . 'J,.; l g* ;*: , ,: .l' s * ' ; 7' l_, g g.! .,y.3.,, 1 ; s,. . . ' . ;,..v";..,..,,,8*,..; g. '! , ,;. . .j ga i,(,,.. - . ' . . , .. .:.t... .. ,<. .;;.: . ?,e;. ..e.,.n. ' :.,:,'.%.'
- - , , ',,e,.,,, .: : :, ,. ." .a . . , .,a<
- - , . r.,.,..,'s.;, ;. , * , _
- 'n:.A ' -. ::, .:..;......'.e:.,
,,s,, ,* ..,.e A--1,;. i *., .... ?'. ,n ;_' -n..,r"~,.,'. .,s,.. it., , ..:.sf: p.i , ; , . ;.s, x.", + ' 4 + a p w. ef -..w e. .a.. s., . C"; :f .,, ., ..,.. (- , .% . ,5 y:.,."....,,j.z.,3.+..,,;J .* .....Vg. 1 , . ... . .... .. 1 ..'.'q', . . , . . ,. ..-.- ' : .'. '. ::s,J .- ..::. .,ls ( n . , ,.l ., , ': ' , -;:',',.(...: ..s ,; . . 3 , .f ' . :; u 3..~.. . , . . . ;, . . s. . , . , .. . - .'..?^
- 5 . . : e. (,, . . . ' . ). -?;*.;..-.,;,:
n..s w' :;
- 9, . G':. l .- v ': q' . , s. .;
. < ,,A ",,'.-;',,- '
- A. ' . . ~ . . '
. f',. ,..4 ; ,,. L.. .".:. ;:" . ' , x y.?' . : , . ,.,' .. : :, *, .. l,zl! . .. ., . .' ) ; ..,Q; . .. s .g.,, ...: v 3, .n, ., ': - ' , '. : .l * ; ,,j ' . ',; ,.j.,, ,y.s. e.f[ ,,,q.,:..; ..,. '.. :,,.,.i..,' g".2,M'.'~',*y.".' . . A .g,.....,r,,.,, K, $- .;/,' ( y,,, e.i - . . . . , . . . , . . ...~.,..'..,(:o,.'., * ,s..P,.,.','. ., . ',4,'*: , ,.a ',.,;; i . . ,: ~ .,. ..} ., . f d ,,., . . ...,;, .. . . '.,,;p , ( n' ..v. .y < ,c ..p . g ,e; ..,..i... . s' g ,,3.;..,. .,. ,, j-,f,'3,,..,f, ,. , 9 ,a . p .*. .,,l,r. , q j. *) .,$ (,.; e ,. . . . - * , .... T , - i ,,4 .j 7,> 'l3 -O. .,.; : . x :;, '.a. . ; *?: . ; g G.*. t . s,' :: .f ;' ; .;'., . zy . ; . .. Z N:.. ,:,,yf5,. ' ,y. ; -. ..,;.' '.e. - c:"..,. ' ;; .' :cy : ;,w <'.. p: ;;, ~: (, , : w .. ::1,. q,'; ;c . _ ,- :, . a: ,. : , . :: 9.1 j ' :- .:. , n . .
- s..,
":'.' * ,,:. . . . % ' a ,*,: .:,. ,'.n.u,
- t. . ?. ., , ,. w. .. . ,,7' . s ..;. u'. . q' .n .~ p,. : . : .- .;- :, . s. ; ; ~ :: .; . ::: ..:
.r'.- . , , ..e,;.',',t.t,-'-s,.Y ,,',:<- l ,?,' :e .'c ' . - e,.*, .. ,. , .., ;. 0:; . ,",
- e. y,"... ,,.g g : . . ~ : . .- m .c .,t:n :L. ,, i;.- ,j ,g 1d , .:. -
,.- s, , . ; .. .h.,. a.m. . ;;.,:.+ . . .e.. 'p.. , s ; a:, . - - -s' ' . ,- ;p 6 '" a , " ' , :;. -' .. *4 - g*.,,. ' ; . ,, i .g,e...... .,.,..,.,o.,:,.. s .- 3- r e,,,. s.- * * . .,
- , . . ,. . . . ':*: , ?; *' V; ;",,-, -..,._"....:- , .. , } . '.r. ; ';**.:" , Qp a . :9 ,.,,-, . ' . . .; n . , * ', , ,.i ., .'%.v
', ; 't_ .. ., , . '.e,,.: ;..*._6.a ,,1,. .- ,* . * ' . , - 4+. ,. ,. '..Pia %,: :n. ., . .."v. s.. o,J. ,.n, f .. v., . y . 4 ;..i
- o. ,, . . . . . .
. , .g .+ ..,,;- .c.+ ,s . g ,. 9,. , .. ..,,.,,.,.3- " - _.;.. < . r s.- .- . , - ,. ,1. s. . ., . e . . .,: , .,. , ; ,9, . ,. - . .t I 'l'*..'.'.' ' . [ .' , Y ',), ', $ ., * .,kM . ;), *. ,[ . P . .' , . 5 k.. ;t .k *. .'d,; .,c['..,'., ,,.. ,' l.:[..h..~.,.',,,.,', \- s .m - ., . .. . ..:. . . .r . ,a : :.,...,,.' m ., . . , v e:.- ;? ,- .w a . . , . ,. q . . *.s 7 a.
- .>,- , j,,, .' , ' y, :b; p. - "= %. . s'- 'o
...,.'(,.;, . a ' t..,:,, ,. . n>, ,.s., ,.. .> ig, f' ", .
- y , , :k . ' . -; . n. ...,.,y,'e.- , , . " <.,f..,.3
[..'..,'",l . , * * , ^(. . . l ,.-l, , L( . . :,,.. r..4. ' - -.V. . ?. ,, *,l .. , . .n., : *.t.. .s g. .. e.. v. .u s; .'l "[" .:' p ;,J f, ;).* ,. .(.i .s h $ ; (.,.:,'..,,;....,,.i ,. :3,.......: , r( .., ..',:' .-.:r,,.,...i '-','.4.,'.(,, .'*.',,n..,'...,... r. s:,:, m- ...*q'.< e, ,; q~,,,,, , * .,=:. 3 ;; ;;. .. 4 m,,,'.'%,. '; l. l-. .. ; ...,,q. * %.,* . .,,,;._'.,-*', .U.*._,.....y,.' .t.' < s '. .A.. 1 - s . * ' . *. r_. 's .s i VL ^. t e. / Y \,, f , .,; ,j "g . ".- . ., , , .; ' , - ' . .,l : ', y ' ,i ', * :' ' , . ,' *. f. . * .,',4%.. (. j .. . .. *), *} ..,'.e.,..,,_ ' , ,., ,<"...,,'.,.'.'.:..e.b. M., ,r,. ".,,r 0. ; 2 . '. , - ,.e.'.. - / i g.. .,;g ,- . ,,N... . . . . ' , u.r,.' s . '- ) i s...,..9,, v,e..i,,...e.,' . ' , . .:.: .,. .;;i .S .-p-6. r . . .':. a .a%.i .g , . . . , .*i..;,,*, i a. . . j' . ..a.,'..- : , , , . ' ., ,- is.,. ::*n * .* , . =9 . \s.q ... ':.,*.*a.,'...,- 7 *) "a*....,, r: ., 4, ,m... . < - ' . . . . . , ,s.',- . , , , , . .. g ,,L,.f.. . ? ? - -'..,,.,.','.?.,i. m - ,- *).- ...r,. ) . , ' .. . i . .; . 4
- e. .l ' ,'s. e. s,,
,., s. - .,?,ity'.,f,..,.Y.,, , g,, . , .. . , . ,. ; - ' . .; . ,.,J',. 3, p".. , , ; , -,,',.,.c,8 .,".,:.'e. ) i, ,; j .lr. ,.* , . . <v.,,.j. ..,1,.p . , J ,.T,*,.l#. ..,,.,'.j/C ,..',,.,.1..,;..'*-.,.,.,.5 f ,...r 7,, . , . , ,, , , . , . . . .,7 ; p ,* , 3 , , . . .. ,,, .....s .:,,.'..A.,,,
- - . a. s :
;*. ,. 3,u,'... ,,t,1 -. , . , .,i+>'.~."c.1 . . ' ' .p,. ,,.. , . , ,.U, 1s .* > . .g.,,...,',....',..., 1 . ..o.,,,g .3 ,,..,..:c.;,;..-gi:.,.... s . .. ,,6,.s,,.. , ,., ' : . k .'a,-.j *,.;. . . . , ;t. '. . . . .r.u,,,,.,t.,.,.,.. *y . .. .; . - c p, , . J' s ,
- , , ,~.. a.i. .g....
- n. A r - .,e> . js +g
., ,. ,...g . , ., r,..,,,'s. ,r..,..,h, 3 *.v..,c+....s,...,$.,..,",...*.;r ,; - .. . - C.... . ,y- ,N* ,.".,..,p. 7 ., -p J. .,,,.,,3.t4...,y, ,,,.,- s,g ' . es ,-s c e. .,i- i,s.. ,',} . :# . ;>vu. ,, ,l-~s...y. , .: ..,,t.,y.,h,.,v,.,. p ?., , ,..s,,-s , (. ,- .: -
- . *- , '.v. t. 3 .
. ; '.^;M ;< o v. 'l.',,',.: a '- .,u- . .i * -a .t; ,, 5., .,)-., -i '.';-, t e .' .l ...: ., .: 7 ',,4...c , , F- .. : s . ' . .r'. s ,e ..f s ; #,,$ 1. . . 4. t g(,' ' ' ' 0 ;- I ',I '.*J," :5 , ';, Y* ;; .. l l.', . ';,:f,2;.' - pel:: :_i , A; ".y .f &, / . , .:':; ;/ll, 7 . * ' . I 't,",,- .i, . ' .? ,[. '. . ; .. ' u,, ','rt,',.'~ :.".[4 '. . . . , s - .. p ;,;-' :s t .}'. ., , l', n ;, ... . %l ,, 6 ' .~;; .:.t.s . ' - 1 .: .. { , s< :* n,;. . . ; ' '. e! '!' ..V* . c ' ',y;..c;
- n,,_,' d. . . 'sy ,'";,
' : s ,' ,. 4lq_.-: .-ll*. g" ;so.,, ., h; *l,, :,. ' .r,<- Al .:l;$lQ - l ',f'-l . , . ., -' r. ,, y ,. . , j. ,. . " , y ,y., . u + , 'f l f' f, ,< _. f. ,s : n.s,gyr,9y*,r .,,.: :W s;^:c).z..(., 1* . . - .
- l.p'll?'
.c . . . . *! '- y ,.x, c'f.
- ,.<; . ,s.
,', . :h.*N :\ f. l,.f.,
- > :s., ;. v :, ' - s,r : ( * :
. ",':..: +c s
- . g]
- i. , ; .'
;.; .;d.%. . , .? : : . % " w .. . , ,, . ; n' , *e.+,- .@,~ , \ , ,; <> ,r....:*,
- r al , _ , 5 ,. ;4 .
- . s,..,y' .: h., y> ;>,, -),e b.y.q.c
- y. .. e , ,, .,
p'*g.,f .,;; .'s..,, ". .; p i.. g,. ,1;,s.,. s. ; ,s .5.. ,5 3. A, 4,f,j c Y ; ' .9. ;"' ..;.-. * , .3 y..- ...,.3........:.,i..a,, y ,c . a.h'..,. 4; .>.p.
- s. . ,
.,g.,<,. T< . ,,f p, y . . t +,,. / + .. ,;i , .. . , s..y..v..- .,'.,J,, . . ,, . , . . , . . . .i , ;. y . . ,..,,4q,.',.. . . - , . ,,,f i. g ,<p,,.; :., ,,4.. . , ~ , ,. .s .4 ... : ),:: y,.;.,,.,.. ' , o,. 1.i., % p ., .i.:.,- le t ,, .i. , ... f,,-r, * ..s.;i - . ;p. .,. . m u., ).g ,,6,,f . ., c.s:g .,n...q,., ..
- w .' :e : ,.' .
. c ', ^ .' , - . : . * . $,l . :.'., .'i , , , . e .l l.f. :';, )* h* h ;d, i. , . , , . f.'.f.-l }kl .Q,p;) .,, yf...'h ;. f G. :^i V. h Y ,Q. y ,,\. ~ , :ll dsyg ,t) k f'.' &:rf' . :'y. ,~ }ki$tf ;: N;; ;. .;t '.. Q:,;;CQf,Q :,Q ,. ",i[,". .2;l*',.e Q' .): y l,. % '!$( Y:f .i. G ;','".il:(A".: :q' .3.':lib.h,R'.I_.b, Qf,;.::' &b ?i 'fI' ';? ,. 7.t
- !pa
.s. [:Q. .p;. . .. { La .,..f '$fE ,'*,t Q,. . ..f, }}y t;. . &, v. . j))$l,.), y,. ,p r .O,.. f,1::( d* .,ig'.},,?l:,Q;f.', l., p:.w, ; ,4 s. a*. e. . .M , n ,9, . [" s@. e d, ':,.1, ..% ty,',p<g .3 )c.y, % ,, .*..Q,'& .;;'..n;j;. M ..'(c, _ ;+..;y <c.
- 6*
.v s . w.:. _ .:p. . tc. w - t * . s . . c4 . ,p;,. . ,. .s),..,,. .. ,c. b ~. ,J.;,,...9...- ; .f.Q). ,.'.[. ;;.y,. l.;,f,; ;.'. Ql.,:;y'. ....t..,..,.. .g ...,. . , ... .+..o. , , p... ;s r...v , o. , v .. _stg.. p. . s, a { , ; >;a, .s.3. :. . ,, ; : ., gtw;e frf ^. , . . o. .e,',h,,; ;y ,g
- g. ;
k,. . 1 . < : 'e f . ... 4 , .. .? ;., -V1 .,f' ',. .:a:?. ;? ".' .' , p l ..',';
- h '? '
ny..y& L %l , yY .6 ,., .' }' 'by
- f. . , .,}s .
?. . .C: . l .: ,l. ,$l h i 'l - + ;r
- b,, h;'l,5:h... f.. .:f:_
l ' v o.f:'h ' k... : ?*' m *- i, ; ; ' ,. ' N.l&l.'d' = %...'.,.\ : I; \ l? .+ + j + .. ?. a i:, . ~;, , :$.0:'.%,D;'.;;. '.r*;1. .ag %'i sy: &.g ..*.k.t 1,,t . ,u~ , ?' p .: .,l'.,,- V 'jl: ).'S- r. f.,;. l.,(; s..!..e'9 , . . "( *4 .v-p s< . s n ;. p1 t;y&'[, ':: ,Y ;i(;0. h ,.:'1lcJ _: ty, , . . lt O! k;g , , 1 ,, i ' . $h f f . '. L ' S ** }l ?. .N'. ','. .. .':,: ., . . : l h:s* ?, 'jh' . . ; fs'l'.\,,L.,f - . ,E .h,, ? N',h*.:?. '.h*f? . ?, . /m. . (.f . , ,',, ,, ~f.0_ , ::'f: h". .n.e. c . 7t,,,f . ,{'?.,1. < h,'.,p,~ s . .I',A, - ~Y. .; , 'v '. Tl . y lL. ?;;h. .,.,'().f.'. ., , R . , .. d . r .. ,S.l, yi . , i' . ,. a .. < :. . : ..S, 4 ,,1 - : ! ..y g .:e ,y ?- ,ls'"; ": ., .3 v.t :;. k ye,, 9,3 .. '.,,;...,'.,'; , , .j; , j,b ;c,< :.
- +,. ..**
..*e..;,.- v '3 ,_'oj? 1 - y. ,,n. ; e q.,c, . -,. y , .- f, , . p + ,,. Y v,,..,,.,',.:,_...,:-, ,; . n .. - ".- - r - .; :. . ; . 2 ,.y:: . > : . . ~. . . > Y"i. . Q:. ,.v"',u.~ .;, .; s .G -m Y., ',tmt; ' u.,.. .,% .7,e.;.9.
- . . i.,ii.. .:. V,;-(p, p 69.,L.. . c .;,;- ;1.
. :: f. , , . c.' , . . , . . , ,. . ': . i . ,; 7 . * . M g....., g I , . . . > . , , . .~ . , . , .. . . . . - . . .. i, .c ... .. . . . . ... . g g ,- .. ...t. , i . .. 1 ....,i .. t . . < > . u . .: * * ,, . [ j ',l' , ' .
- l. '. ,. l ', , '; ,
,' r '... ," . ; .- 1,, . - , if l,, el;, ' '. f, ~ :;i )': '.g, ,':?* ;l n . :& .., E
- li l .*5l. E ,... r,";.
} '. s ;, , Q. 4 ' -?. ' . ' ; . ) *. 'l. ,.Ml).e. , ,a.. >r k V- p. ,\"2 W.f, )' '.g' ....": : : > L ; }.. _ :, . *: .; : . :': '& ; N ';~ ' ; f.y y@> .? :.'.'. fD ' d * :' !; ^'.? c',. l*g'p . ". .: . % . ,., ,l ;C . . " ; r. 3l.l': if'i.js : ~ : ; ., ,, , :' *l ;,,n. , G. ;,I. '. ' , ' .* '.b.kr T:l, l[, . ,;.;' '. , '. j , . : ;I , . i ..i 7y-{. : ;: (: ~Q %.,f.c ...,.,,.v: ' O .l.l,. ", q[h.s. ' ' Q l" l:. ' , ';*,, _ < i ' ; ; . 3.j. ': y ,, d,._ , .. ~.' x, G.,' C '(!;.h., . .... , 'l ? :ll 1. l.. l - ..*. y,. l~ f'_ - . . ' , . p. .m :i . ..- ,Yg<; ; ,.:.,. 5, y Q., y , , , .. .;c......,.. s.- e .t, ; y ,,. g; , , +.. .. ,y ' . , , , -s - c.. . ,' . . , ~ ' . . .y e , e. , .t. ,e .g .,,g.v, . , . ,,.f,.~,,;,'. . . . . .- *.. s ;\ .:..,.,* a . y; m ' ....., ~,'.';',.,:' e . \ ,z,:'i[. q .g...,,y,:,.,,.~.,i p,: , . . . , , . . -. ,a. ; . , ,y.. , ;_ - 5.,- '_.
- r. "-,
,,4.,. ; .",;s .- ,' % ,e i .. , .y u ;i n: :m . i. , . '. * .,. , . . '. je'.g ;. . t .- ,: e ' ' l~ ; ,'- : .' g *e l , ~~. . 3,?. ..: g y ,: ,;r,s_.., ; . ;, . .:<,' 9 ' . ,,'..:. :,. . . , .. : z,, s,. ' . ~ - ..dl'._,..'.',, ~.' .,Q (Y'..,,,p' .,h.'. ' .. .,i :[ 'r. .: ", .} , .I," - N. .".! l ; %. S. ' ' ' Q ': ~., ,., . .; -C', A . ' ., ?y r ' k' ,' .:, . ^ll ; i ., ,"; * . .% . :,'5' -$*:. '~ *?. ?'..!..'' l5 \ :* . l,' } , f; a 9. '. .9 .f.;..- ..,,,' :, [ ) . - ". ','z! ^ f:,/ ( ;y ,.* W .. ;.u. .;r ',- '.<'e c. ;.:,.'.. .. .ecy,.: %, '.q< . .' . : .a . . . ..o,.- . . ': A .T' .; . w. , . ,, ;n .. t,. ' 's.* . . . , ' s'. - : ,...e . . . ,'. . n . .i,s , ;.."',C' ,,&..'-. A ;. c.s. l ".;,X,','....- ;-:ge.e. , v '. ' .e .'r.,,.l.. s, :v. .,si , ', m' .. . - . . . :: . . ' : . . .. . . .1< * ,._ r' 4.s ' 'f ! '9. ;, ' ',..l G,; . " , .',m.,,... ,rn ,','. .' . l.' [ , Q i, '; '; X ,",. ; } ,l ' . _ Q ,h' , _ ..;.g ,r '8'l, N,., ef ',0. ,,' <. . . ', . 3 ., , ,
- s. ** . *' .. - . . . .' .
'- . i .- - s 's. ;, ; .. . T . i' &:,,4r S,,; f. . ;; ...,, \, . ~/, : ' l.. ', < ; ,f' f rk. .f' ,ji ... :.'.,.'.c' ' ; , ' .,, ', ' a'( - 'h f. e ,, * .. ' -'.**.,~.i * * *'..,o',.h',,',a'}. ~ ",[' t *,,f. 7 , ,*s' 'i ' .h . ' , f),- ;'[f)' .- . ' , '8 ,: * ,.b v['o...- 's ,,, ,, c*7., * ,..r,. - [ .'. h y' . sm=,.t.- . . - 6 aJ .'* %.,.'e..p:", .t4.,*,* , < ;i..:.,"..'.a.e.,...,,.., , ,s' - \.+ .j,.'. .'i',,.. . .e .s -.p !',,p.' . ,* h, 'r g" .i ...,n;.... ',' - . ..; .g. s., .i,s' .,,.,,....c.!A.(,.r,, . .. , y n.r. .t..... 9".,-. ,.g.* s, . t e 3 .. .a . }. .o /<. ",,e y'*j,,*.., 6 d). .,I ,,,r.. . , #'.. go ., .,g,.,.w'.g,.,.....,.4' p, , ., [ , . . . f d'j cy,,,..,.,A. .. . .g,...<,, .9 , = ' 'l'y w ",,s, .s.."(.,,.,*3.,,n.' *",i*'.?",.'.. .,,y.,e,* i '9'* e', , , , +f N',.4 p i ",' y. - .,j,, - ' . - *s,,.(,. i '.9 ;.. . l ,p . . . ' , . l... , . , "s, 4, ,,t pt s, - t.,,,'f,a,,.,,.. , . . ...'8 .?,... , , . , e : . . ' , . . , , . . . ' ' , .,',,'i t ,'.",, fe' ; .' - - c , f.n,.',> ,,,' p' , .- - - ,,, $r , t .* "%. , ' , *-j" .... . ,8 .l .l<.,..,'.,'.eg,4,.,t- ,, cj , . i p ,, , ' ' : . 'f *- .,,",. ,L . : , *. . * . . ,. ,,,,,,'.<.4,,/..',,,.p . . - ,:' .'. <' , -'t:*. :. . " .. 'v.,,oi*; . ,.:...?,.',,., .......a.J.....,*;..~.....- s'.*' .',; ',3,,k.:...,'..' %': .- , . .',. ,: f,./.,',de; J . ~,q.. .,P', .,l , a ,( .'s. , ".a. *. . ; . .. s;, ; .> . . ,
- ' :' ..' c, - .,. '"., z..'.'q 1 .' - ?. , ., . . ;, .* .',.:.-l', c. 'r. ' ,. . .- .' :'-(..,. . ; :: \...e.*.
- ,',,.,?J J . . . '.n.o. , , ; ::
'.'..,, .. ri l'r e g. -. . . e s J '..:- .':* .,. s e t, . ,- / .
- l '
s,<o.. .< '.y"':7 ' h;, ' ,. '+-, l.- - gy . .*,,. .
- p .:, ; b*
* [,[ 0.: ';.,; ._l ;. ,f. '.$y.?, .; ,. ),'. .,',\ .','a,k:'h -;f';) * ? 'n (. . ',f. x:- ,.' l,'f f. f..?l.;r5 ~. ^ : ' .. . '. ._ ', . .. '..,. .;a. l<f, ~ .'. ,. ., '._ "., ,. .b ' : ;:tf. ,'?. . . .. l.-,m'lll.Y '.%,. ; .,.. .j'.,m. . ..! ,,'. , .I.... l ;.'.' '.h)', Y .I,f.y .,. > . . . . . ._ . . . . ....s , ,,?. : l*f ,.' l.l.s , . ' . . f,;gl ..,.L., . ; ';. ..v's f.i . . :h. ...'? i , ... . 3 .. . s. 1 ... ....... . . , i , l,1 f .: : .l.. . f, '; ' *';, , .._f . ' [. _ N, l. : ...: :. Y . * , .:, ' . , 'l; .' W
- z. .' *t.:.;[q$.: m. ..:ah,' . *.y:s. ',:. , . w. . ,:n, .f", 'w:f '.;. ."; ',.;l .l .rl:, .. s .yn...e :, :v. w . ..:n. . .,.,.+J., . ,m, .. s:.nw.....
. 6 .- ,, . , e. .., . . . .4.. . w s.. . l m, s.s. n(:,, w,y.,.n m. y :.w. .. g E; *. . '. . m$.:. ..,.,a ' l . ;.((.,h*.'.,',.'.,.. -. n' . v w ..',.' h:[ k,. ~' a .y . f . n' . .g . .-'.'[ . .z. '-:. ' ' , . . , ; ..? 'U. k..'.:,,'.*'..x.'..,. , 7. e > : ,'I ' . ,',. .:.- , . . . ;. ;. .: ?. . -: e. ;. . . u ; . -, . . '. . 8 _ l . ,. . . , . c_, "'t + F'. .. n*. . ~9,y ., c,.: .'.l... t J '^'
- t $
Q::%h' ,'{.. ,1 :; ;;
- l 5.' . .%.';; . :h;'l;; A '.9^. *ll:',.f ' .l.i s
. g. &, ;- ,.N 'f:%':S : l ' . -l '- :: :v. .' .): N ' f G }'. y,p ':. 'G'u,[e h'i )N nf l . &' . : ' (% 'l .. ' _ fj.:.:l, -. *, ::. * ' & ' O.b,'.2 3.". < ' .. :;>.Y.';.y: 1. .%. . $(^hlll.l &;Q,}l,$.iH #' ,yM. V-Q). Q*), W;4. - l.., ; y,.?.;j:i.'.q: ; ; ;;% ( '.. '.),%. i ;,y,:'... %'. . ; .'::,:s: d';n,. ' . ;s'3:m: ,'3f? fe,W G. '. ;' *. .$.& ' 4 .h . . .]. f < 4 - .: .y.. V . s,'.,f. ,,....,;- . . ;, ', ?^ic . -. : p:,;g. ' y,.,y; ~,, ;J . ; :.j,._, F'. ;').,,i y~.. .h ;.: ,h.::W: s , ,, ' .. m : y.nM . i $. ~ q, .,Y< . .:'_?..'. .f',. , :.;'
- y; ,.y
~ . : g s .>.Q:,r.!;;?$ [ y. .:. .l 4 v::;;". . . .]. ..r_ j.f:{ , L 'll3. .k. v.:l,) , .je. i;:(.,I. . .t ...'.: ,h,J'T ;N h~. k::'.'? ll:{qk ' h if."y];..,...p n.: i n ,'Q[ f. ,~T., ' . .. .f. .. ;. .c ;. .ylc. k -:. sh . 'f.a;;.$. l9.:j l ;.Y , n .14, .. n :. ,r; in: g:,;; .h' flh*) _. ' ;. . . . $ 2:n~, a : . c.; ,: . : :, .; ~. p. . 1: .,. y . :): . . .q. . T . Q 2 .. . n .:,3. . . - ,, ,' . ' , > ; .f. . /., ^ ..; l 5 .'- 1 '. [ ;', .' , 2; /, - ['. . . .'; . i , . . -. . . . . < . . - r_..'. . . .! 4Fl: ;- w; .. .:. ,-. -.~O- ;<;f, :. ,f. . . ... :, . -, ~ . , - .9 c.,.n.,.;,- * * . ...s-.. . . . - ;'; '.: . ;'.:l.9 . .. !. .. .- ,. ... .: _.4 -. . . ,e, -.. a .: : . . ._ s .. .;w. . . . , ' , - r ..., - . , .. > . , ,..-a '. . . : c. . . ~ g. ,..e. - s ..?.. . - ;Q. : . . .;;; ,.n' O. ., ... ", .. - .%. :*u :.; . ~ * .. :: .. y- . . . :y - .,e : ) . f , .' ' .f.',.,'. . . . . . . , . . . . .. < .u :-.; ~ y _",,.; x .. ...b,..,,....;. ,'.: . . ' " : .:.g. .. -..,;.,.,,... 1 ;;- .. . . ...' . .e;,,;<f. Q ?. p , .7.; : '.,. -_ ; '.. .':; ' , L; . y . ;'.[ ,, . ' ; ..e ._,:...,.,..','.':.- .. . ',. ,i . , ' . , ' .g. L. .'( q .
- y'.i..'.,
, . .4) ' ~. . . ...' , z. ; . .s,.,.m,.*. e: y :: < . . . 'm^:v .- . ~ . .. - ~ pl _ , ... ,.i . .. . ' s -. . ~ ...- T. c ; ... . ~ . > .r _ s Ls . .. .. J. '. . :V .9.. 7. ... ..;. .$<.g. .. ., .;_ . . .- , . .; .::.' n..,..' . * . . '- . .,'. : 1 . n. ..: . : e.: . . . . - .:^. ,; ;.*. . . .5.. ; .. ;c . ' f : ' . . .- .'. : n.}}