ML20212H549
| ML20212H549 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/18/1999 |
| From: | Phipps M HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20137U375 | List: |
| References | |
| HI-981983, HI-981983-R04, HI-981983-R4, NUDOCS 9906240023 | |
| Download: ML20212H549 (250) | |
Text
'
Holtec Center,555 Lincoln Drive West, Marlton, NI 08053 OLTEC 1NTERNATIONAL TelePh one (609) 797-0900 LICENSING REPORT for STORAGN CAPACITY EXPANSION l
of OYSTER CREEK SPENT FUEL POOL Holtec Report HI-981983 (Non-Proprietary Version)
Report Category: A Prepared for GPU Nuclear, Inc.
Purchase Order No. 0708826 Holtec Project 80334 COMPANY PRIVATE This document version has all proprietary information removed and has replaced those sections, figures, and tables with highlighting and/or notes to designate the removal of such information. This document is to be used only in connection with the performance of work by Holtec International or its designated subcontractors. Reproduction, publication or presentation, in whole or in part, for any other purpose by any party other than the Client is expressly forbidden.
~~
'9906240023 990618 DR ADOCK 050002 9
r "et e a wD e
id v
e n R )a e yl b
a O A-.i c
c, lp
&p Saf i
( .
'. d e
O w N r i e r N e s) v e R wl e a.w t l a n r O
I E
I ieit i r e
.ui e S vi P r d nv I
D eIn( e e a e r V
E O R t b o
dMe l
u R O t cAn o u g
~
S T
N l
V- 8I" t I 1
/
n h $
/
q, %.
L
?.
t S
l E
- f t
d i
e r
u inQs eee rhi at s nq u r
E M 3 t
D e
a Q
t
- 9' 9 r
- y k Y
f u, g
. 1 9
g t
i.
q e
r eie ic s r U 8 T d n
h q h t o
n d nk sor C 7 R e M cs pi wo i
g
/ a e O O r r pv A:
D R
O l
i T
U 8
7 l
d P
E R
G N
I S
G O
L ie R
l e
w v
e a
M
=-
wp
/
N t
C f
- \
h f f d t
a n
e m
u c
o l
l t
a ot a' e l
a oR r a pep ht tr . s n oe s nr A- :. N E E i p
c O A f d e
aCw e e R mgv ni W
I D.
W T I C J i n a 6 t a
L I r l gs ier T R U
M N E
L A
N P
p Y
&A E
S
[f , e r
y
- i t
n e Dow R
O M
U G
I t '
< 0
![ f ,
fe t
a s
s lixh"e e t v
S it F C R r
e s) c dnt h ha O H M e k e e W e
)
E E wl t ei t C ost .p pws t
D a n l ms. C- .
V G W ei P b u D
i it c h f D(
A K
E vir V R I eIn( P M M $ PU k b
\ g A t
m
- o. 2 II i deec oin l
k a e u any t
cm _
T C J v V rh hl -
S I A E R p g aCt p pp a P p L
K _ N D hN 3
((f)f, t " h f r
[& ly /g e r nwi mm po r c coo p;R r
N M
, i C
(([f 0 lytabie er u
h[f $,
i E 1 A t eh t H
C T
A L
R O 'et M " 2 f $ 8 f/
8 d i s ic uf mh oiut v
r y e a y n eb N U l n D a
/q o [4 r.
[4 . eve r
r t h 04 t cl O C p n pi s N H C
O 6 NM c- 6
^
U d M 4 W
I L n I. gad u T A A a 4 a aih s no A
L (4
~ m e C )iiv C r a C
$ /.-
o
- m oD . e s(r erp I
F O ht u z. - -
A -
f f t r
t c a e I
R E
A M d(
a ; M M M d & /.
t
[ h-i t
mniwt ai soi i r t peo u V
a (~; bd rprivd e e hQ h h h h& k r
T N
G I
S R
O cQ M eap v ead nvs e a a, sa c.
na r
r .
E 4 P E
kh al s r e D 3 R r s) t l i
r o c e h t, 3 L ol t ht t an I r S 3 I G & S ewbt r ut at ma on e
i i 0 A C
ui AI nP(
r A N e- R 'W W W W W W E h c u c u nh 8 O T s I
N I
I C c s l C C O C C C A
D id h on w pio b.hr t t t
oa h C 8 D t
ck c i sic -
E D o N u osk dowr r el h T L O. . 6 o o o o. :
A c w n E
. o. o. o.
o - -
2 E r pis oe rh m O o 2 x 4 6 B o 8 t T T i
t
~
s 7-9 t 8
A N kr th becisie t D E of mh or r
o n J a 0r O d d M wonntut t
R E
P t o c o ~ o a
v o
J J o u
o e
o o D
I E a or uf T _
E n s o o o '
G fi e oe -
r l t 1 _
B P e T i n.
i 1 i 1 i T- 7 A A oi s t gk r Y m c T c 3 T. T. T. T- T e e L s vp ano M T u c 1
c e : e c N nerhl _
c A i
U c G C e G G t
= G G, A o ci f w N
T T
N E
D o
g S S s. m s:eS ;
5 G
S t
S S S V O
R M
T iht r ct oai PELBA yn s l
y C M P C _
E U P E J
O C o A J O .
R O N A R P D ' 4 Q P 3 * '*I "
yh s7 h 9@ M, W . $ 5Q 3n7 e. I 0 ,aOx g0@ xUE ! r- ?',a a,$ - jo" y*
rt e 7 ea wD e F id ve a O I
R )s e yl r
ab d ac ni ol O cp ep Saf i
( .
- t. d O e w
N r i e r N
O R
E e
wls ae)t v
e s le we e
I ei i -
ui E H it vir e lnv x V T R
eIn (P -
e r
h ae r E O b wMe o h R U t cAn n o
'S il d e
iqs I
R 8 e
[#f i r
u ece s bi r aInqu B 3 t a d g f e ie M 8 D s s r U T t nkr C
O T
[8 R
O d
a n A t
o n
e A ;iw o
is o D 9 P E
r e r a
- v
- ot _
pm Wc R w t n l e
O
~
R i e w f e l
a i v "l e I
H N G G e f m t T N O R w u c hamR .
U I
S L l a - 8 o t s
- t. s rr A- N 7 d oe 1 D E
E R
p i
c Cs [ )#- de nCw a
eni e n
T L C U i T 8 t a mgv L T I
L T r /- l gsie n U P e ne N A r Dow han M E N y i R M G t
fe i
l ts "e te W O
P U
C I
S r O a s
xh i t v a
O
) O R e V c dei nth h N E wle eian)t e
C V D E
G W itvioPr i p # t os ppwn l t
. t e n K
T D
(
A K
E e l( p A . Hldta ny i
cE O V R I
g u e T
S C
A E 6 e s oin ht po p N
U K
P N
R D
WOiu.nws us po rca c t
a O
D C
B O
I T
N A
O N l yt abi s ic i
e er _
N C A R 'e - f ir ut mhteh y 3 L O p M:
M e ne N
O U
C H
f D e
A f a.p n parf Ve=b t
rs y _
I T L U d a 3 aiga = -
A A A s shE n A
[M C C c o M i )' / -
Na eDe C. (s I
O ft t re N F ht r .
I u c o e .
R E A mn pe i i t t nwtu _
M, aioio s : -
V bdrpeevd 2 N T r eap ri r
G R Q v ead r I
S O
P
[ nvs e a a,a nac E E kh : n -
D R rs) taNr ss eee O _
L et t ht t s i in rewbt i
ehb N A
C uirn AI (P M E hcu at n a c cmht s uton 1
I W T NS N H
C A
D id honoa t
w pio r s b, h t h
E 8
D D ckcsc u r el ii h
M T N doskw S L
n O E A ewroec E 0 o - T T pis r h m i
ka h i, sice Y i
t r 1 A N t t
o D E ef ;. ih tr R
- - P -
k D M w on ..t .tT e 0 E a e E a E
B P Y
m u
a T
o N A G A
fi e oist or oe ul M T e T,
L s vpa gk n ea U c c g A N ne i T o A o rhl ci w l N N D $ V i h y T E O M t t
ahi ct yn s f
D C 6 R T PELDA N E U P C J
C o.
P A
E E O O N J
O G R A R E P D 0 1 -
I 3 4 5 6 7 8 9 1 1
_12_ Q P *8 8 8
- L t
7
_ 0 lR o O O o o o O 0 o b F O
4 A 8 o
- o o ,
- 8 8 o 2 .
Q
- s. l R o O o o o O O o 0 v A o 6
- o o , 8 o o 0
- o
- e. lR o o o o o o o o s
/ A o O
- e o a 8 o
- o R o O O o o o o o o
. l o
7 A o O O o o o o o o o o R o o
- o o o o 8 o
. l F
A 8 o
- o o o o o
- o
~S I'
N 0 O o o o o o o o E 6 o. lR e M 7 o o o e v
U A o o o o o o . i xt a _
C O bo mr -
D .
O e R O O O o o o o O o o d fi ifa r
R N 4 .
l g -
n O iss o l
N A o o o o o o o o o o a ns -
n O ei U I S ir A-I T
I V
E 0
lR o O n o o o o p o o ww es ht an e
W _
L R S -
U O M
A O 0 Q o D 0 0 O 0 o f e oht N mf K ll R
O F o. R o S O o o o 0 0 o o t oi oy s t
T T J l
b a O S O g o dngd N I
L A 0 0 O 0 0 O o an pk i O K or D C
E l
- o. lR O O g g 0 O o o o C
> t o htew l 8 e :
C A o o o o o o 0 o o o irne 3 _
N st h t _
O un pi 0 O I
R O 0 O T e O O 0 O O nh t A . l i i iwr A C
I 2 eg /
F A 8 O g o 0 3 3 0 0 o h n N _
I t i R rt eir E
V N
- R' l
0 O O o 0 Q O 0 0 o l t
nw e
l i n
r 2
_/
G iweg A 0 O O O I
S 0 g O 0 0 o r a E e wa n D 3 ?
n e O 8 C a st r c t
c iMv N Y o R n ee e s t
/
t d
N i d
u wjo er r b y ?
t r
j or r o i
ht ?
gn h e ai hh s e r c yj e :
e d e pr . r r
!t c r o s e ts fia in o m "'
e e ip v a y p t t 1 8 t u
ht c n m ee l l is x ar d P 7 b o e rh mu e it ance pina h ed c or t
e nd
- O N i t
r t t
n fe r
d?
e d ft n
a o isio av e
n nf o j
of 1 fnf f a o can I i d
s e e i
vR s s f re i
l p oor p sc a2 n *" d i e s re T s r s S
/ r oE
/ e op i r r r kr tot ar o
k y u d g h R e e r o c. c t pS o E O
P b
t n )y b
t u
b t
a t
eR C h'.
t u
t cd u ec do u?
i d
s oe r c vh e odP k w o o we t
l rt n t ell a "
w e
na an r a Y T a-- a t r eM onrar r e l' el l m e. s.
. N d rd n l e a l ust oic uo e a w N ia
- .n en po y d e y d w ieA m r u f s a n 0 E
M L m em at w w? d o mar no gi t k pn r
o er o d ns v i t of at a t a w ah y u' "
u Ts a e vQ e
r e A d x d dR is c we hd yh m?
eb pe idn U R i(s e s s iE sf s e as i t ea t t s m a o y t
one l h at D.
I C E ht ts i
ht hS i l ht i a l s mlo m t m c pm bem r
T O ht re ec E. h ld i
r N ll en t r la e nhc cr n mc o i i yC set e u u N D l f f t
ino :
o a h o mh W s'"ucooyd E l e s ul ht ndi f i i i aR oa r r i D
E G Wlc W MN Dm Af o iataIo s Pir n l nI i Ch w Caw N M E U
C 2 3 4
0 G O .
i 4 5 6 7 8 9 1 E D A A A A A A A A A A '8 L i
7 R l o o 2 l
1 l l l 2 ( L F o.
pA o O l
- t l 1 l I 2 I 1 3 o.
R l o 0 z (
2 ( l l 7 o z N A l o 8 1 I 2 l l I L 1
- R
( o 0 z k 2 l ( (
Z o t n A /
o 0 2 1 n e
I I l I t ode e.
R g 0 0 o O O l
l l o 3 'o f
A 0 0 o O o l
[
O l l l
3 o R i o '
- l l
/ o I 8 e.
g A f o
- o O 0 I / I o 8 o
"S T
N ~
R ( o # o o / I l o l o E O o.
M 7 f o o U A O d / I l o I o
C .
O D R 0 0 6 O o O m
e R
O e.
l l 1 l 1 o t s
y O N 4, S H N A
/ o 0 o O o l f l o 1 o A T O Q U I S s A
o o 'c I
I V o R l l
O O / [ l O 3 o T E te L R s l
o W U A O 0 0 O 9 I
O M
l l l f 0 0 I
ht N R
i w
K O
F e. R l
O 0 0 D O 7 o l 0 3 n o
T T 4 i O N
t c
S A l O 0 0 O O l O ]
0 3 n fu I
L O K g d D C
E R l o 0 L l 2
( [
1 t e '
- o. a H 3 l
e r
C A l 0 o 2 I 2
0 I 1 a y
- 3 N
l 0 t O fe a
I T 0 R l
D O 2 $ 2 l l [
2- 6 s a
A C 2 . m r
A
/
I F
I A l o
- 2 l 2 'l I l 2
O e fo N R r E
e p
V 0 R' l
0 o L l 1 l
( { 7, O t o 2 N 1 d
G o e I
S A f 1 l 1 I I l 2 O i
f i
l E a D u r s eh i
s h c o d e? d e e e ?
q O wt st c e i t f o inr t
i s m a N
$ ie n g ih du ht y's h t c
e?n r e a
rt r o i ht o t a
t o
n cm t
nna a dw t
r y a n c v i mo j c e no e d k
I ~
I erP s 4Q I sr s
t s p lur k r
i d p cm s
a e.
but io o padn?
s a p r w e edi u n u e oc n e d,i es t
r mis d t h b
t c
t u
p e
r a
s 1
O N O
i p
c en e
w s e o ldoc i
h c y s u .s t bl a snt(o iu v
b d si s b eyp o s
) e r s
uf o d u
o m
e P
l
_ I at ie v o r w t u he (
s ys? y t
u v icf gure nh ot nfi tn o( e )
s te r
p u s Q T r n t er ela m n S
' R pa e r
s e h a t l s t en t
ab nte mdan iopnd t i
ai n i t
a em r
(eia kr t
c e E I
O d th d d nc u? dit oh w)
. e ai s n e t
s p r n e lut h l uh c eu dop r o?s u t n Y N P n er a
n a r.yori eo ci sj t
. glai mo t k s c w i
c wo c or wr e d
o a rw T r o e a nht di t nrt la l aed r p r ht N ?
r o t ud %kci o tn r a o o) cd e ec csd t ep o a sh i t p
r :
hto m'd . t p iwf een a ht o l
E h pt e d
ai o itsfii t o et r kr e 0 wt cs t t al er a o t yt M ud o t u uu s? n r pd n ca ad . c f ro o a U
at r ei a n u d o oie md r te nd oe oa hV s r e i i ice np r
e er nen . u?
od n
. of w e r
el ap e he f
orove cl avy i
t s ii i
nt h rl efi ot ) .
i o lts ta s g D. C ht u st st n a c lah coir ua t t i ypr n ladn ht alauaimr sung lad ad epa t 1 t l u c ht T D O e msu e e s
? ek p l akS qe o e a n or ce h c sd i N v cn a us v i s v r a er l l nf o l c t l is s t i t v e v al e t ki et r u s e w :
e ov l
a a a o n r oA li a o a e o r au t t er r opp op o D E Hca ec H b l
wi AwQ Wc ci n i e WD Hh t pHo a
Hot uh c( Awa Di n D l
e n
M l
a N U os E
C 1 2 3 4 5 6 7 8 9 0 1 2 e r G O 1 1 1 1 1 1 1 1 1 2 2 2 P E D A A A A A A A A A A A A ' L
_ . 7 R ?. 2 (
o 2 E I I 1 7 t g F o.
r O f A z 2 f o 1 2 2
2
- s. L o 4 R 2 2 (
o z 1 2 2 7 L t 0 e
r A 2 l o L L z z
z
- f. L o
- o. R 2- o [
o 2 2 z 7 2 q 2 g m A , o o z
l z E 2 2 z 2 z d R o o l o o t O o t p 2 z o y
- A o o l o o b o o 2 g g t 0 o R
, o o [ o o " o o o 2 z 7c g
o A o l o O o o o 2 2
o
'S 1
N . R , o o ( 8 o O o o o 2 L o E .
M O 7 o o o U A p 0 O o o o 2 2
. o C .
O -
D O o R p O [ 0 o g b O 1 1 v D R .
O N 6 l
l N A o o l o o o o o 2 2
o I
~
O
_ U I S
A-I I
V R O o l O o N Q Q g g a o T E o. W L R g U O O D O p O
_ M A l O O 0 O 0 N R , K F
O o.
R O o ] O O 2 O C L 1 I 0 T O
T S J O o o g E N I A g O O 1 O 1 0 L O K D C R o o (
o O E O 6 2 [ 7 )c E
I o.
g I :
C g o 3 A
(
0 O , 1 t o N
O I
T R o o ( o (> 2 o O 2 2 Z O A o. A C 2 /
_ I F A o [ o o E o o , 1 o N _
I _
R E
V R o O l 6 O o g 2 7 7 O 2
_ N o. S
_ / T IG A o o l O o o o 2 7 t N 0 .
S E _
E D s M O wc eu t
k s n i
s t e e
?
d E d e R nu N id r o m o ly s e l,rh r ie a s vo r o nis a t a ff ot e u s f I s e
r pr w S iot ai c s n a u r t ue U ds,go i r .
pd e yo f a i f s T r e y t
pif ly a e l e e t
a f g t
n Q aid r n 1
ork i
U pl e r c ht r c e ka e ht mia o ur eit d )y E d d e
_ t a
eo ue r n u cf i l
? P R N
l ot a no o nidt i d b c i n
~ s t
a c cn w f o?d IN t u s v a s gs ms onl a a t ln S no I a luifi O ar er
_ r t nP c o y o c, de a u es r q? o s,innt t
I p
?
s in i
)
et e e isir r e v p e n e T
R e g h c? d i h pa D sadle htetn n?gd ee dh t fh t yi e n oo mes S Y s e
e (s i d S e on n v f r?t ol u N nf r ng E a eh i
r s ab t i O r e
i w Ti e oe Si cc A es ci n a e eie us t hd e i oo nr ct t
a dela t
oa rk L d ny t
Y P a ea t y d es o ra,sqe e t n t a n i moufo ht u i v c A coel r le me ht r rt S nlys edl t T r h c n )t l a x o (o p oa mio isl c n a N N
o d t f i
eF ay N saa yb d m t nd r s i s n let d l a eP A ep r o nn re asi r ui o e
r ule o dt s bi r E m m w els r O ioa ie foiv dC r n a up 0 r ta c ii o o mix o pi led r ee o a I rd s s t
a? ine ppt n a n I r e io R c q M pl e T po n t t r r s t (
o U F F FS f
ur k i v P u c M mn r
t a efoe c d aer usd p ye t r
e? si m r
t uf ut pl a i e )a O lprd ?
D. C o a ymR ptc e ug e d ed es nl t i e n ut pain t uei t hf u p(
n? mu N pynd ol c G a r e e e a e mr e sf oc?
tat n I 2 3 nehi t ie i l T D O _1 e , e e u U ss sb i
ef ht tum e o pm o l a ca etod nc eC S h ad 2 I
_2 2 ei t v
u ae v gt. h hd S a r ri ni e l a iy N 2 2 2 a t o S e d c e t nei v qu h r u h ). r t
u e) h r E t l r
D E A A A l en s r p A r e s h deh tc e e c t pc t ct pe e rl o D e r gd f e
Ar eon l (i I Ah de Wiwa rd o s po r t t N
M t Aad I ad s t i el a n i I sf I ( Ar eanre E U
C 3 4 G O 2 2 .
1 2 3 4 5 6 7 8 1 E D A A B B B B B B B B B C C L
l' 7
- - R # o o l' (' g D t 1 z T.
F se A 0
o o o z .
z z 2 5 e.
R O o O L 0 g O t -
2 1 re -
A o o
- o r 2 2 2 e.
R O o t (> 0 g D 1 1 l
- t m A o 2 o O I 7, 2 2 O
e.
R 2 o o o O 2- o [
f A 0 L o D # o O o 2 o [
R 0 2- 0 o
. O D o l g.
A 0
- t 0 o o o o l
'S T
N o R 0
- 2 0 O o o l 0
E M O 7 4 U A o 2 0 O 6 o e l N.
C .,
t O l g
~
D R O L 0 R
N O e O o Q O O 0 l Me O
l N & A 0 2 O o o o i
T O l M
U g
I S
A- I V .. R 2 a 0 O O o o o o @
I T E L R r n W U
M A g 7- h 0 D h Q p 0 o b s O N
R g M- K o
O F e. R 2 1 O { g O 7 O l
T T 4 . O S A O 2 2 D O O N I
L
[ o 2 O f O
K C R O O O o o C O . .
D E
l l 3
- e. T. t 1
~
2~
C 1 A O o o O 0 3 N o t t L 2 O ,
n I
T e
R o g- 6 Q ,O 1 7 7 .,.
A 1 C
I 2 o . A
/
F I
A 2- O 6 o 1 t t 1 N R
E '
R p O g
- p 7 1 V
N o. o s g- 2
/
G O 7 2-I S
A O o o o 2 E
S I
2 7 7-D S s e d e
Y n r
o e e O t
n lela b t L i f t as ht to N e bt l a e A g e nl d dn r s yn m
e ipml a ewi t
e r a ot t pna v
a? N A
f o d a r
e d
e t
a u b at i
r mna ec r o e. h h d e pa n m, ht s t c g q % d r 1
c u tntnh s e R ey i
d e p? ee
~ u o oir di .
n de n e e n pd up N q e ci re t il sb n o eis t
a t e e. iw a, c lei s e d
a a e s nu ahicem G si yi O r v n a i
O isynd u( s e s ny z is ay I
T ec s
laem op c e i r
t a mlpN
- v. ahle t cai ol lpin ev m b n o ol a s e m R
O f
a r
i r
ehh ee lyht r
tesifi ir isir u r
o I ys et u s
d dt
?
s mr eo h t e
i t
a?
zno i
t i n dng a ch it c S
E P e t
at t eyt o cl yeqc Sc rod
- pnre vnn i
et e e
,y i s e y i oi Y
t n mdh ue e n a v e r a E D d ru ashe i
c :. mse t ei t on nh T
N i
n derah nc i . r yu? c. a osi Fwily s
- rl yo rO sueyf s e pi t c ss,rem a b c -
. l ah r
cl s o u ci yb e i v) w E g? i f ew n sl oe i p olalat in de qu doi"n.
i ct nro i
d s r is v (r i
0 M isd ee i
ct h o g pn en is x n cgton a c D o gnt y dthl a o d oeh n?s ef f" h.-. m e y e?
p r
df is eot p s ee o af ocr t m e ode l n om D..Ci ei shc od n? deonres a ci s t r is O hei rnri ai l ? t el rey t
h ein mrieo na? ut l egr wbo al el e et e ipze v e ev e vb en ot nsH med ont bat)s mbat t
1 ht as id b r O heai eise hi p e a o e e p ^
r eeo T
ND thed hii ia E Teig t
e i eilct a h h a wic h oy e on t l v n a e t
et hnp d o si att n et ec g h s pm iei hl u t ht he hd t
r e p D
f E p g s ep epa h c l il c M ts ns cl t f s e t ad a r e r Awceox e ri nue oec a s ed s cd at Io l l al p s pe p r pn t
r r M I b I r Aisdb i a r l l I e a Aaa is "a Af o Act oih N U E C G O 2 3 4 5 I. 2 3 4 5 6 7 E D C. C C C D D D D D D D D L I
7 L 1 ..
o t 1 q
- R t z 1 7 F t
- l O
M A 2 2 t 2 t o 2 1 2 L 2 6
R L L t t q o 1 1 t t t j
t .
A
/
ll 1 2 z 2 % 2 2 2 2
_
- R
- l t t t t-1 o L t q 1 a os A L E I 2 1 o 2 2 2
2 o
R t o o g o D o o o
- l o - w t
A L o o o g o 0 o o o 2 o
- R O o o o 2 8 o o o
- _ a l A o o o t
- o 0 o o *
- ~
S T o N O o
- 0 o o o 1 E o.lR M O 7 4 o
_ U A O o o 0 o o o 1 C
O
- D R O o R
. l O O O 0 O o o o o S' O N 6
_ l N o o o 0 o o o o o o 2 l
l O A
_ U I S
I A- I V o R Q D o 0 o C O o O 2 T E R
- l W L c O O U
M A 7 Q O 0 p D C D 5 N R
K O
F o. R l 1 O O 0 o O O O o 1 T O
T 4 N
_ L S
I A 7 O O 0 o Q O O O 2 O
_ K . .. D 2 t o 7- 3
_ E C o- R l
t t t 1 c l :
i C 3 3 A 1 1 t 2 o k L 1 t 0 N
O
_ CA I
T L t 1 1 O 2 i 1 t
- e. lR A 1 -
/
I F
I A t t l L 1 o t t 1 b t N R
E V
R L L 2 1 g g t 7 t
- 1 2
N /
l G
A t g 1 2 o t t .
I S t 1 Z E
D ly d e ly f fd d O 8 es c e e
t a
d ri a s, y oe t n
a N ri u y i
v e b s a s u ra q l
o l c rt s nd oi e or t el e e r o t d r L sb e ndt ht p t r p ? un is la u s on d e oa ? oe pa r -v t NN le i
a hl t
i ,
n s
e mu? qs nd di pr b
t us on e
vA S a e
wesn e r
ped r eem dieecn I t a et r n t s a n o heQ M f ?
' lelu E c l
O e s e u f i ed gi? b htetu3 t
s e i c i u s o ot a t e T ni e p) y s ure I
T ntud f ht q e s yl t v f S h a u l
d a r a a I ai lo ot e e E t
R c s sf u e t r mt o ee pC s e i c ht h S nc T a bi r s l s
T mb r dl ee Y O
P U ide u s l u g
f o c
e e ps cl e, cl a d roy t (
u? U imu la f
n o
i l a a c de e d U tnsp P f s, ut n si e ol a P a n s o vf s O i a n T mu l i bh ey t o E mee N
E T impt U n a t i
b e t at t h s s p r t ea t i t r
e ve T ae U d s t i d
e c e my i r s
e N 0 ur u r r f teb O der u ueo e e i
p op ui r i i n nd A M go e s mr pt O tu ur t d f i oo a s U N i
s epn od r
d o d eDmh it op mr oe e
N tpo?
in?
gl e d n
nt u or c c cc ur L
L ut qen D. C G d e a ed c? o? t.
t n c pa c r G uf d ab e oe E de m 1
cige s u od er e man r u s? t C a er T O ol ut s y I , l l S oht s ht ie s ens la tr l
a fu t I S e ni eo t
e S N D ct e yi yfl e ei E p u si f
o E hi D i res u vm hl d r u v e a qu ed r e h s ht d I t a op r p E D r Aai n i v n a e s n a (c er Ae r r r Af o D ts f eq e t a s e s Dae l a o M l e N M l i i d r i r c r c r i l l E
U G C E O .
1 2 3 4 5 . I. 2 3 4 5 . I.
L D E E E E E E F F F F F F G G
7 c [
- o. R o 3 F a
r . O A O 2 2~ o O o 7 -
_ e. R 6 E O o, o g r
r A o E o o O o -
R g O O o O 0 -
e 7 A O O o o O o t e
a 2-D -
- o. R 2 2 L o o -
g A 2 2- Z g O o a R o 2~
o O o g o t o o
~
A 2 O S
T o .
o o N R 2 O o.
U E
M O 7 A o 2 o O o l i
s a
C t p
i O
D .
O e.
R p k 6 O O I n -
R O N s
_ l N o E E o p O i
T O A U
I S
I A.
I V R 7 Q Q g 3 O T E e. W
- L U
R s A g S -
v g 0 O -
M 1
N R K O
o.
R 2
' 1 7 g g T _
F T 4 O -
2 N S
I L
A " 1 7 Q O )
O E
K C R o D d p o O T
A D -
E o. D I
I C
N E
A o 2 o o O D
N A
3 0
1 1
e.
R p D O o g )c S L -
A A A C 2 I T /
I F A O # o o o o I N
I R
N I
E P R 1 o o (p o
. V R C U 2:
N e. O
/ Y G o o o o I
S A E e E S
E D f o
t s
n e.
d n
t t
o o n,
O L
C O
N e u a n .
- e. S c a I n hu t, v i D
_ na ed icer nde le e t
(
1
_ igm? r n vi q i r bare l
l au c S n uq i
N s ri e pb oa scc eeqr u e ;y c cd T lis inro df r O r r ne pc e N e e n r yn r g r h p E he p d i i e o
I pi T t e he t c S laoN M i
eb tek oof an t
R t r l a
/ ii b h o c Etr l r
e o M E O d t t
? i a n Y e s e w t
P nte c er mdefe t s
f i
y fo dei?o O ph e C l
T a a n cot l i n e t t t N yua c e dl l ac mt R :
qn cf a r et d eb umd u 0 cofo t
E i ee epa n i i a s O lidt epxl e f t ii nroo F M bs ainta o ep t a cu?
e of np c r e "p ?
U is s a ut ne pss on E _
u? k i s _
D. C e c o nm qd oh qd ee s ei uI k oCr o t
ah g s n C A
e eit t
htesseiac yR 1
T O csid d d c tg uri di w o ev P aid f
N D e ai ve pi n ai vN s ynig v c S E eo r a nos ru e s
r e p e
r r cl opp aSi o ws S D M Apr r e ir l pid As Apa l l Uht D "o mi I H
N U E T G C E O 2 3 4 5 6 7 E D G G G G G G S
U L -
I l
l Holtec Center. 555 Lincoln Drive West, Mariton, NJ 08053 r OLTEC *P" "* (S *) 7S7~ S Fax (609) 797 - 0909 i
l 1NTERNATlONAL l
QA AND ADMINISTRATIVE INFORMATION LOG
. (To Be Filled in By the Principal Author of the Document and Placed After the Title Page)
I CATEGORY: Generic Document No: Hi-981983 I Project Specific l
Holtec Project No: 80334 In accordance with the Holtec Quality Assurance Manual, and associated Holtec Quality Procedures (HQPs), this document is categorized as a :
Calculation Package
- Technical Report (Per HQP (Per HQP 3.2) 3.2)
(Such as a Licensing report)
Design Criterion Design Document Specification (Per HQP 3.4) (Per HQP 3.4)
- Other (Specify)
1 The formatting of the contents of this document is in accordance with the instructions of HOP 3.2 or 3.4 except as noted below:
l l This document is labelled :
Nonproprietary Holtec Proprietary Privileged Intellectual Property (PIP)
Documents labelled Privileged intellectual Property contains extremely valuable intellectual / commercial property of Holtec Intemational. They can not be released to external organizations or entites without explicit approval of a company corporate officer. The recipient of Holtec's proprietary or Privileged Intellectual Property (PIP) document bears full and undivided responsibility to safeguard it against loss or duplication.
Revisions to the calculation Packages may be made by adding supplements to the document and replacing the
" Table of Contents", the " Review and Certification" page and the " Revision Log".
l l
i e
t 7 n o w
w n
o a ,
4 ,
D t i s
D r
G- % N< I t
o p
u, i
v e
r r
fo
/ (f e
p A n
'5 O 2
' w
/A A R _
- 3 d e f / / / P e i
N & W d N Q e
n R
s e
3( N d "* H H ht e
2L s r
_ N N
o n
2 % \
~
P f.
f e
o t
o s
a 5
2 O ~
n o 3 J
I S * ',
r-t i L C
t I
V E
R ^ ^ A
- 1 S
/s P
!U a
i l-g s
a m
m r
e
- m r
o S A
$1 g F j
f t l {
- e T
N E
N [& f N d & /c A
T l
i ht o
s d
h e
T.
^ T. R W6 e M 4 -
fo r t c
e O
P U
e e j o E
_ C b e
r p . R
_ O l i ec k
E D
W f
_ ho I y t l T f f b R e t
a i
l a ow F O
O ne E u H D 9 q ioiv T P ,
n f. y tu er X T Y 0 n a loe E U T - 1 4 .pm T A
E T
R I
r e
w e
W9 t z f If
$ f.
a o
c vh et gi n E I
T
_ L O P
i v
e
-- f ht e md E P R R
M [u, u E n MiemO I .
.i o
T L
U R N N M4 de v
a s
Tr f a a s F
E B
D M
R -
O I
S I
e dlf~
t p
n 2
- d n
a l
l in ye a a m E
C A
O 4 V t a ' M )(s. r ed L d4 F E D R t
ae P 5 R mtin
~.
G 3 &r do fn 4 sr G" O 0 -M L-i ip ts d O
0%Nf o -
A c L L 8 nn ht J( uo S e ea N u . ht .y t
no N A o O O - d n-. c - I S
I T
A R
E B
%Bc .
bM a p r
hom i
f t
s oi ay I V
E R
s y S C n.
M v W t
ue .
+
_ I w i ah n . F _
fam F U O _
I N e 4w v -
l y
%$ d*
et hr i a e Y
_ T t a
t z s 2
. t o y f rT e R T
s /
R l gt A _
D s-
& /n b e ac C
V Wp-k3 E - 9 / l t v ne M E &
uo on aj o
C i J ' MrP M r
e j V-O o dr U
%' 4we D - l leo t ce wY 9 R S O w iwe 2
l N P e w f it eh "
A A
%. en i t v Pmo r
&=4g e bt i s Y W R n
e f e6 k& s uh hf er B E o.
M%m fm r.mi c
_ t C y S N
I t- h"t - t h bW I V
E N O
- wM fh V is 4t u wt c
di el r
er a T
N E
R J
/ 'o 4 a #se A55 2 t d M y fl 2 d o
_ /
I S
I n
@ s e f.r. %p r o r oht U V C f E As %s r O nk or bw ei O
/
Nc ?./
r R sf -
f f ,
- t. io t
aw l
l i e r
D 8 . /g wga 1
i
$f f/
Ws t n+M /u )A -
i f
io c
rf r
r
. t.- n a
S I
H
/r h- % e
- e T
/ s /u >. <54 M
a 1 ; f Ved b F 4'
' nd m t
/. f t 0 [ f-c O
/ ^ i gee u n :- je L s
e n Mt Msp fL m
n o (N O R
M &~ Db dy ea e
i t
o c
s iP he t
f ht r
R I
E T
e s o y N B t ni
' 2 I 4 " 7 s i t
am r nb O M et - ~- L c s o o o C mn A s s ~
r s N ui o ie e
U W V M W i
~W a t W'
v O W'W-w"'m p
N ct r OE e i a em I oo 8 RT h ic h r S T DP W 5 S PA T- C Abe I V
R O
P E
R g' 2' 3~ * 6' *
~
t s
e o
E R
E H
T
p l
' l i
w
_ e s n t
a u o D p t t
o i
i v
s
& s C 8 p t e
_ r -
- u. r e
b g /g -
fo
_ E i
w e
v g J
/
d /o J
Z VQ -
2 3
P Q
d e
e n
e r o
R n I I
r ht e
e N /,( P o N
O I
M '
f.
f d
o t
s a
n o
5 2
3 S
- e i t L n C I
Mm O a n
y y V ig :
t E * {^ )
. s s n m m r
S R a 3 s4
- i g e F
o _
a t
T g c
- 6 e N T 1 i
s d
h e .
E ht T T R
M A 7 e r t O
U f fo e @-- P E
C p b e
R O i -
eck E D 4$ l i
f y ht lo H T
R e - fi f b F O E t
a d l
a ow ne O D 1 .qu H
T P
Y & y 99 g y n
ioi t
u r v
e T
X S.1 {- l E U T r e ' g a
p oe vh T A T I -
n T
7 b m o
et n E E R e . 1 y c gi n
1 1
L O de e ia r 1 P 4 ht u E M. 2 P d Lu I
T L
E R N o
. R O'
q i
d n
e v
d ere q
e R
O F
E B
U N a ff r a 1q e s a: D M
R O
I S bW y d n ly-l E
C Mf a, a .
I -
O 4 V te /+u g ) i r
A
^
a s ed L F F E D t
ae P F r 3 R )4 U- e mtn
- G Q b, Q. (e 0 &r .
1 w sri G O i v
ip s d O L 8 ht o ~
- er t
nn L
- y e ea S
%m N u N t
&0
_ , ht .y n O -
A 5 y d n om c O I
- I R (o na ap s S I
T E r m i
t
- - V A
B h a o fQ o E C - ,
c yS R M w I
F U & ^o 3 1 q e=t h e n .
am F
O
$ ev f
N s
u =t I
e -. - / q,h t ro ia e Y T T t
a f, y f rT e R R D /1M ( bl e g t A
C t c E
C E & vy 1 h
p t v uo on
=
=
=
MrPo j
e M D
N J
O R O r
e-n wM, %s Y: q.d er o l
t ce M
U S
d h l
w eh A P de e
- d i f e en jt o
A wp%=.
r b s Pmo Y W R a s- ti er W;
s D- B E o dv uhc hf t t S I
V E
R 3
N N
O a%
n o.w co w i *r MR 3
1 A
t-t-
g i t
l t k c mih sw cu yu b p Jla i n
I T
N E
8 I S Asc U {
ed h o ah r
o M 9 "*-
9 I Cp U V
g DE Q t
C s+T, M yg,,
ei
/
E
- nk or bw O 8 R , io r D l C
- t. lj [f e./u.
t l 7 1 l ie iaw wga g c r S _
- ; ' e f iorf . tn n a
I H
I H
ay ' /" ,, e/r. d I,3 e Ved nd b
r e
m e
m n
Mt c
T F
ige u e oj e O
'p e
y~ 9 Db s
e n e n
n o
do iPs r L
O R
R dy i
t c h e t T E o ea fo h N B t nt 4 t n ,
i t
am e
s nb t
y O _
M et L cs ore a o C U mn ui o r. r. & A V
s sa we r
e ia sd N N ct r oo W- S4 6 t E
w E OE ei v t
p a
i v a em r
O I
S T
R O
C Ab e I V
E R
P 1 '
' " n YE AD s
e E _
E
$sT A AN F_r t
R > 7 7 7 X F {D QA t t
i N o H T _
_ e o n
_ t w a t io D o t is p v
& u, er r r e 2 f o
w 3 d i
e P e v Q e e l n
. R l r ht e
_ o e N P o C N O
f.
f d
o t
s a
n o
5 2
3 I e L M S I
V e
t a is n
g it a
n C :
_ E D s im m
r g R &
i g
r e F o
r o t e
t m o l s
d S ht e T s f u i
ht h T N A A er T. R E o t c O E fe jk e _
P E
M K b oc r o R U /- le pl .
E i
e b
_ C f y h we I I _
O e f
t f i T _
f li ove F D a
- t a f u nr O R E D ff q o P y it he T _
O Y & A " ' f n
a lon ut X E
l T D yl i
T T r
e 4 f g A Il
" fS ! f p
m vi ed gr e T
E U R e w d N f o
c I 1 inir qu I
T g
p A O i
v e ht u E e dr e P
E E . R [ n d e R o O L R i
d era P N e t
c s F E
I T N P A
v a
s fe e fa ma B L O d yn D n E U I S
l iade C W
l a,
M 4
_ I V e )
s r t en A .
t V L R 5 E R D a
p a p p p l' p > S D
(
e r
t i ar mp P O
F E & yI' I\
0
- P w e s n is a d
G 0 r g 9 y "
I i v O G 8 ht o d N e r
t s ne e r L
O L A u f e ht .
y tntu o
cn a S
N O
d n
_ N O
R E
8 na ap r
o ocm iS f
tsi g I S
I V
I B ht o . E T yam R A
M ue ah ne F U aT C N e e 3 et ht ro f
it c
O Y
I F
I T
C D t
a 1 4
l c
o b* y lf bt e r
ee gm a
j R
A T E A p A 4 lGA uo v nP M
R J r
f e tl u f 0 9 d r on Mhe a
M E O e s Vg t
C R 3 w y oy 4( g h*
L C le o l
t cm je o U
S P e - ienew f
r -
D Y of i
v b r
Pt u A N R e G t sh is ep Y A o
. - u ic ht n i B
N 7- r. mh yl S W t is w b ar I T
E N do N
t l
k c eh N I
V O 2 cu ed r
ei t E
E R
9 I
S I
e 0
1,/n k 7 h or Cp dr w ore M
U N
V E
R D t
a J k s h + *h ds - ior n k t o l eg b a n
C O
A 4 A '4 awr l a D
~
r y J
"* n u D i
f io c iwM t
S I
I o J g ' A rf r nt c I f"w g e d I N ht k N - k k B Ve b e e mjeo T A
u
- f. E' M" B f
- nd igee s n m
u n
ur cP o
F O
s e n d he L pd- -
Db e o s t O R
f e i t ihy R E
d y e a c e f t b e
T t
am s od N B t nt 1 2 3 4 5 7 8 ics r na O et L o om C M mn r
e r
e r
e r
e r
e r
e r
e & A or s e r i ui o e s U t p
t p
t p
t p
t p
t p
t p ) V s a w t p ivb e N N ct r a a a a a a a e O oo h h h h h h h E OE ei h ve h a el rl I S
T DP C C C C C C C E RT T r C Aw i
R P
I T AD e s
E E AN A t to I I
R QA D
7' 1 2 3 4. 5 6 8 t t N T
e o n
_ t w o a t D t o is i
p v e
& u, r
r
_ r o e 2 f w 3 d
_ e P e i
e v
e Q
I n
-/ o
. R I r
e ht e
N P o t
5 Mr N f.
f o
s a 2 .
O d e io n 1 _
M I S
I t e
a n
g t
a n
1 C -
_ V D is i m
E s m r o
9 i r
/ R & g t e F A r lo e t / o s d
e S ht i T S u ht h T N A e T. R N
_ r t O 6 o c E fe jk e P M /( b e
oc r o pl E
R U l-l i
e b E _
C f f
y h we t
I I
T -
O e li a
f i ove F D t R E D
a 4 f9 -
u q
nr o e O
T O
P Y & i S-t f.
y n
itut h X H
T T
T +
r e
w d Mf -
a f
f a
p m
o lonvi ed gr e E
T E
H U R e S c ini A
E O
P i
v e Mf f f6 ht e r qu du re T
E R
L E o R n # i n d er e a O P R N d e
t c s F E
I T N VeW 2 v a
s fe e f
a m B
a D _
L O (f d lyn E h-n U 4 I S i a.
la d e C -
M 3 I V e t
a E D 6
)
(
s ir t e n ta ir A
L F
R 8 E D A N 4 j r mp P e W M./r aj e R -
O F
0 0
r d go Mto t l ie w
v s n is a d G O
b s L o e t ne G f r _
ht
&ag e r S _
uo e tntu -
O u -
ht . N L :
A .
0 d n y
co na g
O I
N O
R E
G. g M na ap rm iS f
t si S I
V ho oc B o . E _
I ya m R .
T M t
ue ah ne F A U et f aT O
_ C N e
. h r t of it r c Y h ej e I t a y l R
_ F I T C D bt ve go ar A
_ T E A ' 9 uo nP a e M R & A A 4 % on Mh M E
J O e r f 4 N he f b- '
d r leo t cm t
U C R $ w N
ay f -
\
l i w e jo e S P ie t f
en of r A
_ D v e 1- - b s Pt u r .
N R q t i sh ep Y B
_ A o
. u c mih ht in yl S N
kr b ar I
_ W is t
w d o T E N d l
k c t
eh r N cu E I
V E
N
/
O I
S Nlla h o ed Cp r dr w ei or t
M U
R I
V e U n eg e C M.B t
k N E R D a
A a A f
p -
it oro a
c w l
b a l
n a
O D
~ & 4 ) i f r iwM S I r o N g N & Q irfo r t
nt c e
l l
i M h t Ve ed b e
mjeo T F
u nd m ur A ge u cP o O ie s n n d he L e n O
- Dbe i o is t h y R R d y e a t
c e
t b T E 0 t
s f
od e N B t nt 9 1 1
1 2
1 iam r o na O et L c s or om C M mn r
e r
e r
e r
e & A s e r e is e N U ui o V sa we t t t t p p p p ) tp ivb O N ct r oo h a
h a
h a
h a E OE he ei v h a el rl I S
T DP C C C C E RT T r C Aw i
I V
R h PA E CE PD O I AD s R P T e E E
0 4
2 A AN t t
o l i
c .,
I l
l l
SUMMARY
OF REVISIONS ;
Revision 0: Initial issue. I Revision 1: This revision incorporated client comments that were provided in GPU Nuclear letter E550-98-070 (dated October 19,1998).
l Revision 2: This revision incorporated client comments that were transmitted by e-mail to Mr.
l C. Bullard on May 3 and 4,1999 (Email
Subject:
"Holtec Rpt. HI- 981983 R1 Licensing Report, Part 1"; "Holtec Rpt. HI- 981983 R1 Licensing Report, Part 2"; "Holtec Rpt. HI- 981983 R1 Licensing Report, Part 3"). The changes are i restricted to Chapters 6 and 8. l Revision 3: This revision incorporated client comments that were transmitted by fax to Mr. C.
Bullard on June 9,1999. Chapter 8 was also revised to include the cask drop l load. Typographical errors on pages 5-4 and 5-16 were also corrected.
I Revision 4: This revision incorporated client comments that were transmitted by fax to Mr. C. I Bullard on June 15,1999. These comments only affected Chapter 12 and the Table of Contents.
l-i i
l 1
l ,
l l !
i SilADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec International i Report HI-981983
i TABLE OF CONTENTS SECTION PAGE NO.
1.0 INTRODUCTION
1.1 References for Sec tion 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2.0 HIGH DENSITY S PENT FUEL RACKS . . . . . . . . . . . . . .. . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 j 2.1 General Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.2 Summary of Principal Design Criteria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 2-2 2.3 Applicable Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.4 Quality Assurance Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 10 2.5 Mec hanical Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 1 2.6 Rac k Fabric at i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2.6.1 Fabrication Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 1 2.6.2 Oyster Creek Rack Module . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 12 3.0 ~ MATERIAL AND HEAVY LOAD CONSIDERATIONS . .... .. ... ................. .. . ... 3-1 3.1- In t rod uc t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.2 S truc t ural M at erials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -
3.3 Neutron Absorber Materi als . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.4 Compatibility with Coolant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3.5 Heavy Load Considerations for the Proposed Reracking Operation ............ ..... 3-3 3.6 Re ferenc es for S ec t i o n 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -
4.0 CRITICALITY S AFETY ANALYS IS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 4.1 I n trod u c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.2 S ummary and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.3 Abnormal and Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4.4 Inp ut P arameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6
'4.4.1 Design Fuel Assembly Specifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.4.2 Storage Rack Cell Specifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.5 Analytical Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 4.5.1 Computer Codes and Benchmarking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 4.5.2 CAS MO4 Validation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-8 4.5.3 Oadolinia E ffects and B urnup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-8 4.6 Criticality Analyses and Tolerance Variations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 10 4.6.1 Nominal Design Case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 10 4.6.2 Uncertainties Due to Manufacturin g Tolerances . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 1 4.6.2.1 Boron Loading Variation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 1 4.6.2.2 Boral Width Tolerance Variation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 1 4.6.2.3 Storage Cell Lattice Pitch Variation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 12 4.6.2.4 Stainless Steel Thickness Tolerances . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-12 4.6.2.5 Fuel Enrichment and Density Variation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 12 4.6.2.6 Zirconium F low Channel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 12 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational ii Report HI-981983
e i
i t
t l
l TABLE OF CONTENTS .
SECTION PAGE NO.
4.6.3 Uncertainty in Depletion Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13 l
4.7 Abnormal and Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 14 4.7.1 Temperature and Water Density E ffec ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-14 4.7.2 Abnormal Location o f a Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-14 4.7.3 Eccentric Fuel Assembly Positioning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-14 l 4.7.4 Zirconium Fuel Channel Distortion . . . . .. . . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . ... 4-15 l
4.7.5 Dropped Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 15 4.7.6 - Fuel Rack Lateral Movement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-15 l 4.8 Re ferences for Section 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 17 l
Section 4 - Appendix A: Benchmark Calculations 25 pages 5.0 TH ERM AL-HYDRAULIC CONSID ERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 I n t rod uc ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 1
- 5.2 Spent Fuel Pool and Cooling System Descriptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 i 5.3 Decay Heat Load Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -4 5.4 Di sc harge S c enario s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -5 l 5.5 B ulk Pool Temperatures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -6 l 5.6 Local Pool Water Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....................5-10 5.6.1 Basis.....................................................................................5-10 5.6.2 Local Temperature Evaluation Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 10 5.7 Claddin g Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 14 5.8 Results......................................................................................5-15 l 5.8.1 B ulk Pool Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 16 5.8.2 Ti m e- t o - B o i l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 1 6 l
5.8.3 Local Water and Cladding Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 17 l 5.9 Reactor B uilding HVAC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 18 5.10 Re ferences for Section 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 18 l
l 6.0 STRUCTURAL /S EIS MIC CONS ID E RATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 l 6.1 In trod uctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 l 6.2 Overview of Rack Stmetural Analysis Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 l 6.2.1 Background o f Analysis Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 6.3 Description 'o f Rac ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6.4 Synthetic Time-Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6.5 - WPMR M ethodolo gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.5.1 Model Details for Spent Fuel Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.5.1.1 Assump tions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.5.1.2 E lement Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 8 6.5.2 Fluid Co upling E ffec t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 6.5.2.1 Multi-Body Fluid Coupling Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International iii Repon HI-981983 l l
t
TABLE OF CONTENTS 1
1 i SECTION PAGE NO.
6.5.3 Stiffness Element Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-i 6.5.4 Coefficients of Friction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 12 6.5.5 ' Governing Equations o f Motion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-12 6.6 S'.ructural Evaluation of Spent Fuel Rack Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 13 6.6.1 Kinematic and Stress Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . 6-13
'6.6.2 S tress Limit Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 14 6.6.3 Dimensionless Stress Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6- 17 i 6.6.4 Loads and Loading Combinations for Spent Fuel Racks.................. ...... 6-17 6.7 Parametric Simulations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 19 l 6.8 Time Histo ry Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 6.8.1 Rac k Displac em ents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 6.8.2 Pedestal Vertical Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-22 6.8.3 Pedestal Friction Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-22 6.8.4 Rac k Impac t Lo ads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-22 6.8.4.1 Rack-to-Rack Impacts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 22 6.8.4.2 Rack-to-Wall Impacts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-23 6.8.4.3 Fuel-to-Cell Wall Impact Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-23 6.9
{
Rack Structural Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-24 l 6.9.1 Rack Dimensionless Stress Factors for Level B and D Loadings............... 6-24 6.9.1.1 Rac k Pedestal S tress Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-24 6.9.1.2 Rack Cell S tress Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-24 6.9.2 Pedestal Thread Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-25 6.9.3 Local S tresses Due to Impacts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-25 6.9.4 Assessment o f Rack Fatigue Margin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-26 6.9.5 We ld S tresses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 8 l 6.9.6 Bearing Pad Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-29 6.9.7 Level A Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 0 6.10 Hydrodynamic Loads on Pool Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-31 6.I1 C o nc l u s io ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.12 References for Section 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 2 7.0 FUE L HANDLING AND CONSTRUCTION ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.1 Introd uc tio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 7.2 Fuel Handling Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 1 7.2.1 Desc ri pt ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 1 7.2.2 Mathem atical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-3 7.2.3 Results.....................................................................................7-4 7.2.3.1 S hallow Drop Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7.2.3.2 Deep Drop E vents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7.3 RackDrop.....................................................................................7-5 l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International iv Report HI-981983 L-
TABLE OF CONTENTS SECTION PAGE NO.
7.4 Co nc l us i o ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.5 References for S ec tion 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.0 S PENT FUE L POOL STRUCTURAL EVAL UATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 1 8.1 Introd uc t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.2 Description o f Pool S tructure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2 8.3 Applicable Lo adin gs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.4 An aly s i s M od e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 -4 8.4.1 Finite Element Model for Mechanical Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4 8.4.2 Thermal Moment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... ........ 8-5 8.5 Lo ad Co mb i nations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 -6 8.6 Input Data for Mechanical and Thermal Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-7 8.7 Results........................................................................................8-7 8.8 Poo l L i n er . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.9 Closure.................................................................. ...... ........... 8-8 8.10 Re ferenc es for S ec tion 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-9 9.0 RADIOLOGICAL EV ALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9- 1 9.1 Fuel Handling Accident. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................................9-1 9.1.1 Assumptions and Source Term Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9- 1 9.1.2 Results.................................................................................9-3 9.2 S o l id Radwast e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9-3 .......
9.3 Gaseou s Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..........................................9-4 9.4 Personnel Ex posures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-4 9.5 Anticipated Exposure During Re-racking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-4 9.6 Re ferences for Section 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-5 10.0 BORAL SURVEILLANCE PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . 10- 1 10.1 Purpose.....................................................................................10-1 10.2 Coupon S urveillance Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 10.2.1 Coupon Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 10.2.2 Surveillance Coupon Testing Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-3 10.2.3 Measurement Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-3 10.2.4 Surveillance Coupon Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-4 l
10.3 In-Service Inspection (Blackness Tests) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-5 l
10.4 References for Section 10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-6 11.0 IN STAL LATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l - 1 11.1 Introd uct i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 - 1 11.2 Rac k Arran gement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 - 3
)
11.3 Pool S urvey and Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 -3 11.4 Pool Cooling and Puri fication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l-4 1
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International y Report HI-981983 1
TABLE OF CONTENTS SECTION PAGE NO.
I 1.4.1 Poo l Coo li n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . I 1 -4
- 1 1.4.2 P uri fication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l -4 11.5 F ue l S hu fili ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 -4 11.6 Installation of New Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l -5 11.7 Safety, Radiation Protection, and AL ARA Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l-6 ,
1 1. 7.1 S a fet y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 -6 1 1.7.2 Radiation Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 -6 1 1. 7. 3 AL ARA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 -6 11.8 Radwaste M aterial Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l -7 12.0 ENVIRONM ENTAL COST / BENEFIT AS SESS MENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12- 1 12.1 Introd uc tion n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12- 1 12.2 Imperative for Reracking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12- 1 12.3 Appraisal o f Alternative Options . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1 12.4 Co st Esti m ate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .....................12-5 12.5 Reso urc e Commitm ent . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-6 12.6 Environmental Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-6 12.7 Re ferences for S ec tion 12 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-7 SIIADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec International vi Report 111-981983
LIST OF TABLES l
l-1 Fuel Discharge Schedule for Oyster Creek Nuclear Generating Station 2-1 Geometric and Physical Data for New Fuel Racks 2-2 Module Data for Oyster Creek Spent Fuel Racks 1
3-1 Boral Experience List - PWR Plants (2 pages) 3-2 Boral Experience List - BWR Plants (2 pages) 3-3 1100 Alloy Aluminum Physical Characteristics 3-4 Chemical Composition - Aluminum (1100 Alloy) 3-5 Chemical Composition and Physical Properties of Boron Carbide 3-6 Heavy Load Handling Compliance Matrix (NUREG-0612) 4.2.1 Summary of Criticality Safety Analyses 4.3.1 Reactivity Effects of Abnormal and Accident Conditions 4.4.1 Fuel Assembly Design Specifications 4.5.1 Comparison of Calculational Methods @ 3.1% Enrichment 4.6.1 Reactivity Uncertainties due to Manufacturing Tolerances 4.7.1 Effect of Temperature and Void on Calculated Reactivity of Storage Rack 5.1.1 Partial Listing of Rerack Application Using Similar Methods of Thermal-Hydraulic Analysis (2 pages) 5.3.1 Oyster Creek Historical and Projected Fuel Discharge Data 5.4.1 Summary of Key Thermal Hydraulic Design Parameters 5.4.2 Oyster Creek Fuel Design Data 5.8.1 Results of Bulk Pool Transient Evaluation 5.8.2 Results of Loss-of-Cooling I 6-1 Partial Listing of Fuel Rack Applications Using DYNARACK (2 pages) 6-2 Rack Material Properties (200*F) (ASME - Section II, Part D) 6-3 Degrees of Freedom 6-4 (DYNARACK) Numbering System for Gap Elements and Friction Elements (2 pages) 6-5 Comparison of Bounding Calculated Loads / Stresses vs. Code Allowables at Impact Locations and at Welds 71 SummarypfIn wt Velocities 7-2 Material Paper.es 7-3 Structural Definitions ofImpactor and Target 8.3.1 Applicable Loadings on the OCNGS Pool Slab 8.7.1 Controlling Safety Factors and Load Cases t
i SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International vii Report HI 981983
LIST OF TABLES 9-1 Results of Origen Calculations for Radionuclides ofIodine, Krypton, and Xenon at 100-Hours Cooling Time 9-2 Radionuclide Properties Used in the Fuel Handling Accident Analysis 9-3 Data and Assumptions for the Evaluation of the Fuel Handling Accident 9-4 Preliminary Estimate of Person-Rem Exposures for Rack Installation 10-1 Coupon Measurement Schedule SHADED TEXT CONTAINS PROPRIETARY INFORMATION llottec International viii Report HI-981983
L p,
LIST OF FIGURES
! l-1 New Fuel Rack Layout 2-1 Pictorial View of Typical Oyster Creek Rack Module 2-2 Seam Welding Precision Formed Channels
~2-3 A Cross Sectional View of an Array of Storage Locations 2-4 Three Cells in Elevation View l
2-5 Sheathing Shown Installed on the Box l 2-6 ' Adjustable Support l 4.2.1 Correlation of the k-infinite in the Standard Cold Core Geometry with the k-infmite in the i Storage Rack
! 4.2.2 Limiting k-infinite in the Standard Cold Core Geometry for Fuel Enrichments Between
! 3.2% and 4.6%
4.4.1 - Cross-Section of Typical Storage Cell (Calculational Model) 4.5.1 Variation in the Rack Reactivity for Several Illustrative Examples of Gd O Loadings 2 3 4.5.2 Correlation of k-infinite in the Standard Cold Core Geomety and k-infinite in the Storage l Rack l 4.5.3 Limiting Correlation of k-infinite in the SCCG and k-infinite in the Storage Rack (with Examples of Gd203Loadings Superimposed) j 5.4.1 Oyster Creek Spent Fuel Pool Discharge Scenario Cases (i) and (ii) 5.4.2 Oyster Creek Spent Fuel Pool Discharge Scenario Case (iii) 5..*.1 ~ Spent Fuel Pool Cooling Model
- 5.8.1 Bulk Pool Transient Temperature Plot for Partial Core Discharge 5.8.2 Bulk Pool Transient Temperature Plot for Normal Discharge Scenario 5.8.3 Bulk Pool Transient Temperature Plot for Full Core Discharge Scenario 5.8.4 Fuel Pool Time Dependent Decay Heat Load for Case (i) and Case (ii) Partial Core Offload Scenarios 5.8.5 Fuel Pool Time Dependent Decay Heat Load for Case (iii) Full Core Offload Scenario 5.8.6 Post Loss of Forced Cooling Pool Depth vs. Time Plots 5.8.7 Oyster Creek Fuel Pool Temperature Contours 5.8.8 Oyster Creek Fuel Pool Local Velocity Field f 6-1 Rack Numbering Scheme for WPMR Analysis j 6-2 Oyster Creek Acceleration Time History - OBE Event, X Direction j 6-3 Oyster Creek Acceleration Time History - OBE Event, Y Direction 6-4 Oyster Creek Acceleration Time History - OBE Event, Z Direction 6-5 East-West EQ File 6-6 North-South Earthquake Plot 6-7 Vertical Component EQ 6-8 Schematic of the Dynamic Model of a Single Rack Module Used in DYNARACK 6-9 Fuel-to-Rack Gap / Impact Elements at Level of Rattling Mass 6-10 Two Dimensional View of the Spring-Mass Simulation SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Internationai ix Report HI-981983
r
+
, LIST OF FIGURES l
t I 6-11 Rack Degrees-of-Freedom for X-Z (and Y-Z) Plane Bending l- with Shear and Bending Spring 6-12 Rack Periphery Gap / Impact Elements 6-13 Oyster Creek Nuclear Generating Station - Vertical Pedestal Force vs. Time ;
- Run No.1, Rack M, Pedestal 3
_6-14 Rack Fatigue Model 7-1 " Shallow Drop" Scenario 7-2 " Deep Drop" Scenario l 7-3 Isometric View of the " Shallow Drop" Model
! 7-4 Von Mises Stress Distribution in Storage Cells 7-5 Plan View of Post-Impact Geometry 7-6 Isometric View of" Deep Drop" Model l 7-7 Maximum Displacement of Rack Baseplate 7-8 Elevation View of" Deep Drop" Model (Above Pedestal) 7-9 Z-Dir Stress Distribution in Concrete l
8.2.1 Plan View- Reactor Pool Area
.. 8.2.2 ,Section Through Central N-S Girder 8.2.3 Section Through Central E-W Girder 8.4.1 Finite Element Discretization of Slab 8.4.2 Finite Element Discretization of Walls and Girders 8.4.3 Node Designations for Slab and Girders 8.4.4 Boundary Conditions l
11-1 Oyster Creek Rack Installation Sequence (Step 1)
Il-2 Oyster Creek Rack Installation Sequence (Step 2)
- 11-3 Oyster Creek Rack Installation Sequence (Step 3) j- 11-4 Oyster Creek Rack Installation Sequence (Step 4) l I l
l I
l l
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International x Report HI 981983
p l
1.0 INTRODUCTION
The Oyster Creek Nuclear Generating Station (OCNGS) is located in Forked River, New Jersey near the Atlantic Ocean. The site is approximately nine miles south of Toms River, New Jersey y and fifty miles east of Philadelphia, Pennsylvania. The station is owned by Jersey Central Power
& Light Company, and it is operated by GPU Nuclear of Parsippany, New Jersey. The reactor l building at OCNGS' houses a boiling water mactor (BWR) which was designed and built by General Electric Company. The reactor core contains a total of 560 fuel assemblies. The net electrical output of the plant is approximately 640 MWe. The plant has been in operation since December 1%9. There is one spent fuel pool at OCNGS, which is located between elevations l 75'-3" and 119'-3" within the reactor building. The spent fuel pool is a reinforced concrete structure with a stainless steel liner.
l The fuel discharge schedule for OCNGS is shown in Table 1-1. The racks that am presently in the Oyster Creek spent fuel pool provide a total of 2,645 fuel storage cells. With 2,420 fuel assemblies currently stored in those locations, the plant has already lost its ability to offload the entire reactor core. For this reason, GPU Nuclear has contracted Holtec International to design and supply new high-density fuel racks to augment the storage capacity of the Oyster Creek spent >
fuel pool. This license amendment seeks USNRC approval to install four additional racks l between the north wall of the pool, the cask drop protection system and existing boraflex storage i racks. A plan view of the new fuel rack array is shown in Figure 1-1. The total storage capacity l
with the new racks is 3,035 fuel assemblies. This supplemental rerack restores the full core offload capability at the plant.
l The new Holtec racks are free-standing and self-supponing. The principal construction materials for the new racks are SA240-Type 304L stainless steel sheet and plate stock, and SA564-630 (precipitation hardened stainless steel) for the adjustable support spindles. The only non-stainless material utilized in the rack is the neutron absorber material, which is a boron carbide aluminum cermet manufactured under a U.S. patent and sold under the brand name Boralm by AAR Advanced Structures of Livonia, Michigan, l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 1-1 Repon HI 981983 i
The new Holtec racks are designed to the stress limits of, and analyzed in accordance with,Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code.
The analysis, material procurement, and fabrication of the rack modules conform to 10CFR50 Appendix B requirements.
The rack design and analysis methods employed in the Oyster Creek storage capacity expansion are a direct evolution of previous rerack license applications. This Licensing Report documents the design and analyses performed to demonstrate that the new Holtec racks meet all goveming requirements of the applicable codes and standards, in particular, "OT Position for Review and >
Acceptance of Spent Fuel Storage and Handling Applications," USNRC (1978) and the 1979 Addendum thereto [1.0.1].
Sections 2 and 3 of this report provide an abstract of the design and material information for the new racks.
The criticality safety analysis requires that the neutron multiplication factor for the stored fuel array be bounded by the kort limit of 0.95 under assumptions of 95% probability and 95%
confidence. The criticality safety analysis provided in Section 4 sets the requirements on the I Boral panel length and the B areal density for the new high-density racks.
Thermal-hydraulic consideration requires that fuel cladding will not fail due to excessive thermal stress, and that the steady state pool bulk temperature will remain within the limits prescribed for the spent fuel pool to satisfy the pool structural strength, operational, and regulatory
. requirements. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 5.
Demonstrations of seismic and structural adequacy are presented in Section 6. The analysis shows that the primary stresses in the rack module structure will remain below the ASME B&PV Code (Subsection NF) [1.0.2] allowables. The structural qualification also includes analytical demonstration that the suberiticality of the stored fuel will be maintained under all postulated SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 1-2 Report HI-981983
accident scenarios in the Oyster Creek Final Safety Analysis Repon (FSAR). The structural consequences of these postulated accidents are evaluated and pn:sented in Section 7 of this repon.
I l
The structural analyses to demonstrate the adequacy of the spent fuel pool reinforced concrete structure are presented in Section 8 of this report. A synopsis of the geometry of the Oyster Creek pool stmeture is also presented in Section 8. i The radiological considerations are documented in Section 9, and Section 10 presents the In-l Service Boral Surveillance Program l
Section iI discusses all aspects related to the on-site field work. The many steps and
. requirements in completing the installation of the new racks are detailed in this section.
Additionally, one possible rack installation sequence is presented in this section.
Finally, Section 12 presents a cost / benefit and environmental assessment to establish the prudence of GPU Nuclear's decision to proceed with wet storage expansion.
All computer programs utilized to perform the analyses documented in this licensing report are benchmarked and verified. These programs have been utilized by Holtec International in numerous rerack applications over the past decade.
The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety in n:spect to all considerations of safety specified in the USNRC OT Position Paper, namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.
SIIADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 1-3 Report HI-981983 L.
I 1.1 References for Section 1
[1.0.1] USNRC Ietter to All Power Reactor Licensees transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978, and Addendum dated January 18,1979.
[1.0.2] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF and Appendices (1995).
l l
l l
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 1-4 Report HI-981983
. 3 8
9 1
_ 8
_ e 9-g 1 1
m 1 o ) tr S
t k o c p g a r e n e R iyR nt i i
a c r m a pfe
- t _
'7 e a 5 .
R C (A 1 6
2 4 -
e g
m ) .
N t o k c N _
O a r O I
I g e T T n y R A _
A t e M
_ T i c o r R
RS mpe af O
OG e C( aB 9 0
5 "
2 7 F
_ FN 4 2 3 - N _
EI I _
L T Y _
UA R
l _
DR o A e o l
EE uP T HN E E F n )
. I 1
C G f i s R 1
S ose ie P
- l O
e ER .i ol b
m R 5-GA b l
b a N m e s
P1 T R E l a e . s a S AL t os s 0 8 0 2 8 6 4 1
0 4
0 8
8 4
6 8
0 6
7 8
4 6
3 0
2 8
0
)
s ie0 N
HC TA 4 6 0 8 9 4 7 4 0 2 3 5 7 8 0 2 4 6 l 0
I A
C SNU 2
1 3 3 4 5 6 8 1 1 1 1 1 1 2 2 2 2 - b 2
- m (= T I e N DK s h O
_ s s ct
_ F i e a a C L F l b 0 b T ER m 6
5 g n X U
F C e
(
e r e i
l E
_ R s s o u T cf D E A r re E t ol s l ed ue t
c a a
D e m Y F rg r A
_ O f a l r on H oh c l S
fu a oi s 6 8 2 8 8 7 6 8 2 e 3 4 4 8 4 8 a g 4 3 4 2 1 6 2 6 5 0 8 7 4 6 7 8 8 8 ND 2 1 1 7 1 5 1 1 1 2 1 1 1 1 1 1 1 1 -
d r a a l
oh c f
f i s
e od t
a o o t t D y y n t t
i i l l i i w 5 2 b b o 1 3 4 5 D 7 8 3 6 8 1 9 4 8 0 2 a a d t 0 2 6 8 n 3 7
/
3 n
9 7
2 7
/
3 n
6 0 8 8 /
2 8
/
8
/
0 9
/
6
/
8 2
9
/
0 6
9 9
/
6 0
/
2 0 0
/ e e ht h l a
u 1 i 1 1 2 / 2 1
/
5 1 1
1 3 1 / 1
/
4 2 2 2 s s t
n h 2 e o s se
/ / / / / / / / / / 1 / / / / /
/ /
S 9 5 4 4 3 1 4 9 1 2 4 9 2 1 9 9 9 9 9 o o i
t l l a t t n na m e
ap lp o l t n
N e e I e h h c l
TT e c l t
y A B 0 1 2 3 4 5 6 7 8 o C 1 I 2 3 4 5 6 7 8 9 1 1 1 1 1 1 1 1 I * *
- I I
_ 3 8
9 1
2 I
H
. l l
' : ,e l Ei
~
i _ = K C
_ = A R .
~S
. g G N
I T
S EI E ~ P. _ =
I X
E
_ =
I Uh.I ~ C.
F e
. Ig. li r.
~ _li ll T
_I0 U 1
I J!1 !i K
C
= -
O n A = -
I R
E Y O - . oO r G A r
i (i
o o
_ o )
(
N I
T S
@ E@ L I
T I
T t
U a E = -
K I
. 3 "4
/ ii . !'
= -
C nE isf f 3 o o -l o Q A
- f f
U 0 -
o0 _ O R
- s. S s 0 - o0 _ O a" :
sS i A o o _ u _
M. l L
4 ,!i t
E i ' i i
- t-o r
iii ! . ; I , l , I l l l E
\A r u- U
- - y~ !
J [-s. = - , F
~
l _ o o - O -
- Oi- - nis ie
= -
i3 O _. _. oo - r k i k s th i I ! . ;1 , i , 1 : :
W jiI su e
l I
l r r rr C A
E lle. e ' s . K R N gs7't l
C A = G -
R N
- = -
e j
K C
A R
G@
I T
N I
T S
I I
1 1
S E I
G I N
I T
S I
@ E =
=
E R
X E U G
" u81 a# -l d ,, :: l.
I F
= .
=
K K C C A A R R G
E N = .
I I
T S
@ I T
S I
@ =
I D E
~ =
=
!i N l ' s I i ' - . '
- _
l e
n
E 1 L )
I^
2.0 MIGH DENSITY SPENT FUEL RACKS 2.1. General D=stion l In its fully implemented configuration, the Oyster Creek pool will contain fourteen racks, four of which will utilize Boral, in the spent fuel pool with a total cell count of 3,035 fuel storage cells.
The four new racks will provide Oyster Creek with the ability to store 390 spent fuel assemblies.
The existing ten racks, which utilize Boraflex as a poison, provide storage for 2,645 spent fuel assemblies. The general description of the racks provided herein specifically addresses the new Boral racks only.
(
All storage rack arrays will consist of freestanding modules, made from Type 304L austenitic stainless steel containing prismatic storage cells interconnected through longitudinal welds. A panel of Boral cermet containing a high areal loading of the B isotope provides appropriate neutron attenuation between adjacent storage cells. Figure 2-1 provides a schematic of the typical storage module proposed for Oyster Creek. Data on the cross sectional dimensions, gross l weight and cell count for each rack' module in the spent fuel pool are presented in Table 2-1.
The Oyster Creek spent fuel pool storage capacity will be maximized using high density storage i
modules that are of the so-called non-flux-trap design. The baseplates on all rack modules extend out beyond the rack module wall such that the contiguous edges of the plates act to set a !
geometric separation between the facing cells in the modules. !
Each new rack module is supported by legs which are remotely adjustable. Thus, the racks can be leveled, and the top of the racks can easily be made co-planar with each other. The rack i module support legs are engineered to accommodate undulations in the surface of the pool floor.
A bearing pad interposed between the rack pedestals and the pool liner serves to diffuse the dead weight of the loaded racks into the reinforced concrete stmeture of the pool slab.
l l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 2-1 Report HI-981983 L
F The overall design of the Oyster Creek racks is similar to those presently in service at many other nuclear ' plants. Among these plants are Kuosheng of Taiwan Power Company, James A.
Fitzpatrick of New York Power Authority, Zion of Commonwealth Edison, and Duane Amold of Iowa Electric and Light. Altogether, there are thousands of storage cells of the Oyster Creek design that have been provided by Holtec International to various nuclear plants amund the world. l 2.2 Summary of Princinal Desien Criteria
! The key design criteria for the new Oyster Creek spent fuel racks are set forth in the classical l
l USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978 as modified by amendment dated January 18,1979.
The individual sections of this report expound on the specific design bases derived fmm the
'above-mentioned "OT Position Paper." Nevertheless, a brief summary of the design bases for i the Oyster Creek racks is provided below.
]
l
- a. Disoosition: All new rack modules are required to be free-standing.
- b. Kinematic Stability: All free-standing modules must be kinematically stable i l (against tipping or overturning) if a seismic event (which is 150% of the l
! postulated OBE or 110% of the postulated SSE) is imposed on any module. l l
- c. Structural Comoliance: All primary stresses in the rack modules must satisfy the
- limits postulated in Section III, Subsection NF of the 1995 ASME Boiler and l Pressure Vessel Code.
l l
- d. Thermal-Hvdraulic Comoliance: The spatial average bulk pool temperature is required to remain under 125*F in the wake of a nonnal refueling.
In addition to the limitations on the bulk pool temperature, the local water temperature in the Oyster Creek pool must remain below the boiling temperature coincident with local hydraulic pressure conditions.
- e. Criticality Comoliance: The maximum calculated reactivity of the storage rack is i such that the true Kefy shall be less than 0.95 with a 95% probability at a 95%
confidence level for normal and accident conditions.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 22 Report HI 981983
l r l
i
- f. Radiolonical comoliance: The reracking of Oyster Creek must not lead to l .
violation of the off-site dose limits, or adversely affect the area dose environment l as set forth in the OCNGS FSAR. The radiological implications of the installation of the new racks also need to be ascertained and deemed to be acceptable.
l g. Pool Stmeture: The ability of the reinforced concrete structure to satisfy the load combinations set forth in NUREG-0800, SRP 3.8.4 must be demonstrated.
- h. Rack Cvelic Stress Fatiane: In addition to satisfying the primary stress criteria of l
Subsection NF, the alternating local stresses in the rack structure during a seismic ,
event are also required to be sufficiently bounded such that the " cumulative '
damage factor" due to one SSE and twenty OBE events does not exceed 1.0.
- i. Liner Intenrity: The integrity of the liner under cyclic in-plane loading during a seismic event must be demonstrated.
l
- j. Beanng Pads: The bearing pads must be sufficiently thick such that the pressure on the liner continues to satisfy the ACI limits during and after a design basis seismic event. I I
- k. Accident Events: In the event of a load drop (e.g., uncontrolled lowering of a fuel assembly), it is necessary to demonstrate that the subcriticality of the rack structure is not compromised. l l
- 1. Construction Events: The field construction services required to be carried out for executing the reracking must be demonstrated to be within the " state of proven art."
l l The foregoing design bases are further articulated in Sections 4 through 9 of this licensing report.
2.3 Anolicable Codes and Standards The fabrication of the rack modules is performed under a strict quality assurance system suitable i for 10CFR50 Appendix B manufacturing.
The following codes, standards and practices are used as applicable for the design, constmetion, and assembly of the Oyster Creek fuel storage racks. Additional specific references related to ,
detailed analyses are given in each section.
i SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 2-3 Report 111-981983 L
r.
- a. Desien Codes (1) AISC Manual of Steel Construction, Ninth Edition.
(2) ANSI N210-1976, " Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations"(contains guidelines for fuel rack design).
(3) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code Section III,1995 Edition; ASME Section V,1995 edition; ASME Section VIII,1995 Edition; ASME Section IX,1995 Edition; and ASME Section XI,1995 Edition.
(4) ASNT-TC-1A, American Society for Nondestructive Testing (Recommended Practice for Personnel Qualifications), June 1980.
(5) American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI318-63) and (ACI318-71).
\
(6) Code Requirements for Nuclear Safety Related Concrete Structures, ACl349-85/ACI349R-85 and ACl349.lR-80.
1 (7) ASME NQA-1-1989, Quality Assurance Program Requirements for l
Nuclear Facilities. 1 1
(8) ASME NQA-2-1989, Quality Assurance Requirements for Nuclear Facility Applications.
(9) ASME Y14.5M-1994, Dimensioning and Tolerancing (revision and redesignation of ANSI Y14.5M-1982)
(10) ACI Detailing Manual,1980.
- b. Mearial Codes - Standards of ASIhi (1) E165-95, Liquid Penetrant Examination.
(2) A240/A240M-97a, Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Pressure ]
Vessels.
(3) A262-93a, Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel. ,
4 SHADED TEXT CONTAINS PROPRIETARY INFORMATION 1 Holtec International 2-4 Report HI-981983
(4) A276-97, Stainless Steel Bars and Shapes.
(5) A479/A479M-97a, Stainless Steel Bars and Shapes for Use in Boilers and Other Pressure Vessels. i (6) A564/A564M-95, Hot-Rolled and Cold-Finished Age-Hardening Stainless Steel Bars and Shapes.
i (7) C750-80, Nuclear-Grade Boron Carbide Powder.
(8) A380-96, Cleaning, Descaling, and Passivation of Stainless Steel Parts, Equipment, and Systems.
(9) C992-89, Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.
(10) E3-95, Preparation of Metallographic Specimens.
(11) E190-92, Guided Bend Test for Ductility of Welds.
(12) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section II-Parts A and C,1995 Edition.
1 (13) NCA3800 - Metallic Material Manufacturer's and Material Supplier's Quality System Program.
- c. Weldina Codes: ASME Boiler and Pressure Vessel Code,Section IX - Welding and Brazing Qualifications,1995 Edition.
- d. Ouality Assurance. Cleanliness. Packauine. Shionine. Receivine. Storage. and Handline Reauirements (1) ANSI N45.2.1-1980, Cleaning of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants.
(2) ANSI N45.2.2-1972, Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase).
(3) ANSI N45.2.6-1978, Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).
(4) ANSI N45.2.8-1975, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Plants.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-5 Report HI-981983
I (5) ANSI N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants.
1
.(6) ANSI N45.2.12-1977, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.
(7) ANSI N45.2.13-1976, Quality Assurance Requirements for Control of i Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123). "
(8) ANSI N45.2.15-18, Hoisting, Rigging, and Transporting of Items for Nuclear Power Plants.
(9) ANSI N45.2.23-1978, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1.146).
(10) ASME Boiler and Pressure Vessel,Section V, Nondestructive Examination,1995 Edition.
(11) ANSI N16.9-1975, Validation of Calculation Methods for Nuclear Criticality Safety.
- e. Governina NRC Desien Documents (1) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and the modifications to ;
this document ofJanuary 18,1979. l (2) NUREG 0612, " Control of Heavy Loads at Nuclear Power Plants",
USNRC, Washington, D.C., July 1980.
- f. Other ANSI Standards (not listed in the orecedind (1) ANSI /ANS 8.1-1983 (R1988), Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.
(2) ANSI /ANS 8.17-1984 (R1997), Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors.
(3) ANSI 45.2-1971, Quality Assurance Program Requirements for Nuclear Facilities.
(4) ANSI N45.2.9-1974, Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Hohec International 2-6 Report HI-981983
(5) ANSI N45.2.10-1973, Quality Assurance Terms and Definitions.
(6) ANSI /ANS 57.2-1983, Requirements for Light Water Reactor Spent Fuel Storage Facilities.
(7) ANSI N14.6-1993, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4,500 kg) or More.
(8) ASME N626.3-1993, Qualification and Duties of Specialized Professional Engineers.
- g. Code of Federal Reentations (1) 10CFR20, Standards for Protection Against Radiation,1997 Edition.
(2) 10CFR21, Reporting of Defects and Non-compliance,1997 Edition.
(3) 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, 1997 Edition.
(4) 10CFR50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,1997 Edition.
(5) 10CFR61, Licensing Requirements for Land Disposal of Radioactive Material,1997 Edition.
(6) 10CFR71, Packaging and Transportation of Radioactive Material,1997 Edition.
- h. Regulatory Guides (1) RG 1.13, Spent Fuel Storage Facility Design Basis (Revision 2 Proposed). '
(2) RG 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors.
(3) RG 1.28 (ANSI N45.2), Quality Assurance Program Requirements .
(4) RG 1.29, Seismic Design Classification (Rev. 3).
(5) RG 1.31, Control of Ferrite Content in Stainless Steel Weld Material.
l l
i i
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 27 Report 111-981983
m (6) RG 1.38 (ANSI N45.2.2), Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants.
(7) RG 1.44, Control of the Use of Sensitized Stainless Steel.
(8) RG 1.58 (ANSI N45.2.6), Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel.
(9) RG 1.61,' Damping Values for Seismic Design of Nuclear Power Plants, Rev. O,1973.
(10) RG 1.64 (ANSI N45.2.11), Quality Assurance Requirements for the Design of Nuclear Power Plants.
(11) RG 1.71, Welder Qualifications for Areas of Limited Accessibility.
(12) RG 1.74 (ANSI N45.2.10), Quality Assurance Terms and Definitions.
(13) RG 1.85, Materials Code Case Acceptability - ASME Section 3, Div.1.
(14) RG 1.88 (ANSI N45.2.9), Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records.
(15) RG 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis.
(16) RG 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components.
(17) RG 1.123 (ANSI N45.2.13), Quality Assurance Requirements for Control ofProcurement ofItems and Services for Nuclear Power Plants.
(18) RG 1.124, Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, Revision 1,1978.
(19) RG 3.4, Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities.
(20) RG 3.41, Validation of Calculational Methods for Nuclear Criticality Safety, Revision 1,1977.
(21). RG 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June 1993.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION 1 Holtec Intemational 28 Report 111-981983
I 1
0 l
I (22) RG 8.8, Information Relative to Ensuring that Occupational Radiation ,
Exposure at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA). i j (23) IE Infonnation Notice 83 Fuel Binding Caused by Fuel Rack l Defonnation.
1 I
- i. Branch Technical Position l (1) CPB 9.1-1, Criticality in Fuel Storage Facilities.
l (2) ASB 9-2, Residual Decay Energy for Light-Water Reactors for Long-Term Cooling.
l j. Standard Review Plan l (1) SRP 3.2.1, Seismic Classification. l (2) SRP 3.2.2, System Quality Group Classification. 4 (3) SRP 3.7.1, Seismic Design Parameters.
l (4) SRP 3.7.2, Seismic System Analysis.
1 (5) SRP 3.7.3, Seismic Subsystem Analysis. !
(6) SRP 3.8.4, Other Seismic Category I Structures (including Appendix D),
Technical Position on Spent Fuel Rack.
(7) SRP 3.8.5, Foundations for Seismic Category I Structures, Revision 1, 1981.
(8) SRP 9.1.2, Spent Fuel Storage, Revision 3,1981.
(9) SRP 9.1.3, Spent Fuel Pool Cooling and Cleanup System.
(10) SRP 9.1.4, Light Load Handling System.
- (11) SRP 9.1.5, Heavy Load Handling System.
(12) SRP 15.7.4, Radiological Consequences of Fuel Handling Accidents,
- k. AWS Standards (1) AWS DI.1-94, Standard for Steel - Structural Welding Code.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 29 Report HI-981983
r-l (2) AWS D1.3-98, Structure Welding Code - Sheet Steel.
l (3) AWS D9.1-90, Sheet Metal Welding Code.
l (4) AWS A2.4-98, Symbols for Welding, Brazing, and Nondestructive l Examination.
j (5) AWS A3.0-94, Standard Welding Terms and Definitions.
(6) AWS AS.12/AS.12M-98, Tungsten and Tungsten Alloy Electrodes for Arc Welding and Cutting.
(7) AWS QCl-96, AWS Certification of Welding Inspectors.
2.4 Ouality Assurance Program l
The governing quality assurance requirements for fabrication of the Oyster Creek spent fuel racks are enunciated in 10CFR50 Appendix B. The quality assurance program for design of the Oyster Creek racks are described in Holtec's Nuclear Quality Assurance Manual, which has been reviewed and approved by GPU Nuclear. This program is des ped te provide a flexible but i highly controlled system for the design, analysis and licensie of customized components in i accordance with various codes, specifications, and regulatory requirements.
i l
The manufacturing of the racks will be carried out by Holtec's designated manufacturer (U.S.
Tool & Die, Inc.). The Quality Assurance System enforced on the manufacturer's shop floor shall provide for all controls necessary to fulfill all quality assurance requirements with sufficient j simplicity to make it functional on a day-to-day basis. UST&D has manufactured high density racks for over 60 nuclear plants around the world. UST&D has been audited by the industry group NUPIC, and the QA branch of NMSS with most satisfactory results.
The Quality Assurance System that will be used by Holtec to install the racks is also controlled l by the Holtec Nuclear Quality Assurance Manual and by OCNGS site-specific requirements, l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-10 Report HI-981983 j
I I
I 2.5 Mechanical Desian The rack modules are designed as cellular structures such that each fuel cell has a prismatic square opening with lateral support and a flat horizontal' bearing surface. The basic {
1 characteristics of the Oyster Creek spent fuel racks are summarized in Table 2-2.
A central objective in the design of the new rack modules is to maximize their stmetural rigidity while minimizing their inertial mass. Accordingly, the Oyster Cmek modules have been designed to simulate multi-flange beam structures. The multiple flanges are fonned from the numerous cell walls in the rack cross-sectional array. These cells are connected through intennittent welds. The weld lengths, location, and size were chosen during the original design of this rack style / series to ensure adequate strength and to adjust the natural frequency of the rack modules to avoid resonance. In general, this effort has resulted in excellent detuning characteristics with respect to the applicable seismic events.
2.6 Rack Fabrication This subsection provides an item-by-item description of the anatomy of the Oyster Creek rack modules in the context of the fab'ication r method. The object is to provide a self-contained description of rack module construction and to enable an independent appraisal of tlm adequacy of the design.
2.6.1 Fabrication Requirements There are four basic requirements for the Oyster Creek high density storage racks. The requirements are:
- i. The rack modules are fabricated in such a manner that there is DQ weld splatter on the storage cell surfaces which could come in contact with the fuel assembly.
ii. The storage locations are constructed so that redundant flow paths for the coolant are available.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 2-11 Report HI-981983
iii. The storage cells are connected to each other by austenitic stainless steel comer welds which lead to a honeycomb lattice construction. The extent of welding is selected to "detune" the racks from the seismic input motion (OBE and SSE).
iv. The fabrication process involves operational sequences which permit immediate i i verification by the inspection staff. !
- 2.6.2 Oyster Creek Rack Module There are five significant components (discussed below) of the storage racks: (a) the composite box subassembly (b) the baseplate, (c) the neutron absorber material, (d) the sheathing, and (e)
! the support legs.
1
- a. Comoosite box subassembiv: The rack module manufacturing begins with fabrication of the " box." The boxes are fabricated from two precision formed l channels by seam welding in a machine equipped with copper chill bars and l pneumatic clamps to minimize distortion due to welding heat input. The l minimum weld penetration is 80% of the box metal gage. The inside dimension of the BWR box (or cell) is 5.9305 inches. This process results in a square box as l
'shown in Figure 2-2. A metal sheathing is then attached to each side of the box, l and the poison material is installed in the sheathing cavity.
The square cross section box with Boral panels affixed to its external surfaces is referred to as the " composite box assembly." Each composite box has at least two one inch diameter lateral holes punched near its bottom edge to provide auxiliary flow holes. For those cells located over support legs, four flow holes are required to compensate for the loss of the baseplate flow holes described below.
The composite boxes are arranged in a checkerboard array and welded edge-to-edge to form an assemblage of storage cell locations, as shown in Figure 2-3.
Filler panels and comer angles are welded to the edges of boxes at the outside boundary of the rack to complete the formation of the peripheral cells. The inter-box welding and pitch adjustment are accomplished by small longitudinal connectors. The connectors are sized and placed to ensure that the inside cell dimension for developed boxes is maintained after inclusion of any reductions from the sheathing. This assemblage of box assemblies results in a honeycomb structure with axial, flexural and torsional rigidity which depend on the extent of inter-cell welding. It can be seen from Figure 2-3 that all four comers of each L interior box are connected to the adjacent boxes, which creates a well-defined path
( for " shear flow."
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-12 Report HI-981983
f' l l
- b. Baseplate: A 1 inch thick. baseplate provides a continuous horizontal surface for ;
supporting the fuel assemblies. The baseplate has 3.625 inch diameter holes at each cell location, except at lift locations. At the four lift locations, a 5 inch by 2 ;
- l. inch rectangular cut-out is centered over the circular flow hole to allow insertion and engagement of the lifting rig. The location of all baseplate holes coincide with I the cell centerlines. The baseplate is attached to the base of the cell assemblage by l
- fillet welds and extends horizontally approximately 1/4 inch beyond the periphery l
- j. of the rack cells. Refer to Figure 2-4. !
t
- c. Neutron aberber m*~ial: As mentioned in the preceding section, Boral is used l as the neutron absorber material. Each storage cell side is equipped with one l
integral Boral sheet (poison material). Only one Boral sheet is required between adjacent cells containing fuel.
j d. Sheathing: As described earlier, a metal sheathing is fastened to the exterior of I each box wall. The design objective calls for attaching Boral tightly on the box surface. This is accomplished by die forming the intemal and external boral
! sheathings to provide end flares with smooth edges, as shown in Figure 2-5. The '
l flanges of the sheathing are welded to the box using skip welds and spot welds.
The sheathings serve to locate and position the poison sheet accurately and to preclude its movement under seismic conditions. The sheathing also provides a l vented enclosure for the Boral.
l
! e. Sunoort leas: All support legs are the adjustable type as shown in Figure 2-6. The l top (female threaded) portion is made of austenitic stainless steel. The bottom l (male threaded) part is made of 17:4 Ph series stainless steel. Each support leg is ;
equipped with a readily accessible socket to enable remote leveling of the rack after its placement in the pool. The support legs are located at the centerlines of cells to ensure accessibility of the leveling tool through the 3.625 inch diameter flow hole in the baseplate.
! The assembly of the rack modules is carried out by welding the composite boxes in a vertical '
fixture with the precision fabricated baseplate serving as the bottom positioner.
j SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-13 Report HI-981983 a
r 3 _
e 8 9
. P 1 _
l s 8 _
l 9-e 1 C 1 1
f t r
ok o NR oa . c
% 9 9 8 8
7 0
1 R p
e -
i h
t g _
e W _
_ g -
i n
p 0 0 0 0 S i p) 0 8 8 0 -
K hb I 7, 9, 9, 7, C S( 8 8 7 9 A N
_ R O I
)
L n n T E i
( i o A U s n
t c
e M
R F
o i r
O W i s
n D 3 2
9 3 2
3 F E e 2 5 N N m W - 9 2 3 5
2 9
7 9
I Y
R i E 4 5 4 7 R O D A F e p T A o E I
1 T
l e n R
- v o P 2 A n i t
O4 l
e D E c e R1 -
L e r b
a l i
7 7 P2 T A u D 4 4 1
4 1
4 S C
I d
o S- 6 6 5 5 N I
S 3 3 7 7 A Y M N 7 7 6 6 T H N P O D n C o T _
N i
t X A c e E C i r T -
I R D D _
E T W D E - 3 A M E 8 9 8 1 H
O S E
G n l
s i o
l t e c e
C i r
f o D o
S-2 2 1 1 N N 1 1 1 1 l
a n
o i
t a
n r
D. t e
I e I n
l u c d t e
o l o
M L M N P I I
l l
Table 2-2 MODULE DATA FOR OYSTER CREEK SPENT FUEL RACKS l
Storage cell inside dimension (nominal) 5.9305 in Cell pitch (nominal) 6.106 in Storage cell height (above the baseplate) 169 in Baseplate thickness 1 in Baseplate hole size 3.625 in Support leg height (nominal) 6.6875 in Support leg type Remotely adjustable legs Number of support pedestals per rack 4 (Rack P - 5)
Remote lifting and handling provisions Yes l Poison material Boral*
Poison length 146 in Poison width 5 in (internal) 4.5 in (external) l l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-15 Report HI 981983 i
F
+- +
i i lii i i
- N '
s f
% /
% /
N /
l I
l l :,
kg p FIGURE 2-1; PICTORIAL VIEW OF TYPICAL OYSTER CREEK RACK MODULE lHI-981983
+ +
L
+ +
3 _
8 _
9 _
1 _
8 _
9 I _
I H _
S L
E N
N A
H C _
D _
M A E _
E S M D
R L
E O _
W F _
N O
I S
I C _
E _
R P
G N
I
) D L
E
(
W M
A E
S YE)
RLL ADA 2 IHC -
L I IWP XOY 2
ULT AF( E R
U G
I F
1
+ + l FORME 0 CELL BOX CELL
\ /
1 I I i 1 1 i i l i i I ,
I --
1 i i l I i l I i l 1 1
'I
_, l l
l 1 4 1 1 1 i i i i i i I
1 l i l I i 1 I l 1 I I i
l FIGURE 2-3; A CROSS SECTIONAL l VIEW OF AN ARRAY OF STORAGE LOCATIONS 1111 - 981983
+ +
+ +
CELL
- PITCH ~
ff . ll. .
ll __it i n I _ _ l l
i 3 i I i i i !' ,
(
l c i , i l 1 i I
, l 1 1 l
i
' , i I i 1
- .s_ ; , _ .;
ACTIVE I POISDN LENGTH
- .s- x -% - x..
l l l RACK CELL j LENGTH l
)
, i i i i t 3
' i i i l i il l I l l l
1 :: +
l l e l l - -
l
..9- - , . . . . .
l
~. .- t
\ i / , , u , 1 I T , , , , , , q t
BASE PLATE FIGURE 2-4; THREE CELLS IN ELEVATION VIEW lHI-981983
+ +
i i i
+ +
t m
b j
l
' I i
O CQ Ex.1 l
Z :
O 9 Q E %
% 4 j W
^ A ;
E-CO Z
Z it: !
O
=
0 Z
N E~
fJ Z
CD en b
I N/ i b
l t
i FJ
+l +
L i
l M
, e
.m
.e CTb l I l
I w E b - w b
A
\ Y 1 u h\\\\\\\\\\
a v/A___ ~
- ..M-m N
Wk # e wDW) /
R_.
9- P- J i
WI @ Q pg /_
s _/_ :
3 <
W . b a k,..
dy CQ Dl .
5, *- y G
s e
- h l l ? cv 1 E~ j l -u 3 y I
I \-
3 e -
{
H 's l
t lll
n 3.0 MATERIAL AND HEAVY LOAD CONSIDERATIONS l 3.1 Introduction 1
1 Safe storage of nuclear fuel in the Oyster Creek pool requires that the materials utilized in the rack fabrication be of proven durability and be compatible with the pool water environment. This l section provides a synopsis of the considerations with regard to long-term service life and short-term construction safety.
3.2 Structural Materials The following structural materials are utilized in the fabrication of the new spent fuel racks:
- a. ASME SA240-304L for all sheet metal stock l
- b. Intemally threaded support legs: ASME SA240-304L
- c. Extemally threaded support spindle: ASME SA564-630 precipitation hardened ,
stainless steel (heat treated to 1100*F)
- d. Weld material- per the following ASME specification: SFA 5.9 ER308L 3.3 Neutron Absorber Materials in addition to the structural and non-structural stainless material, the racks employ Boral ,a patented product of AAR Manufacturing, as the neutron absorber material. A brief description of Boral, and its pool experience list follows.
Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum.
Boron carbide is a compound having a high boron content in a physically stable and chemically inert form. The 1100 alloy aluminum is a lightweight metal with high tensile strength, which is protected from corrosion by a highly resistant oxide film. The two materials, boron carbide and i
- aluminum, are chemically compatible and ideally suited for long-term use in the radiative, thermal and chemical environment of a nuclear reactor or a spent fuel pool. Boral has been shown [3.3.1] to be superior to attemative materials previously used as neutron absorbers in storage racks.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 3-1 Report HI-981983 l
F Boral's use in spent fuel pools as the neutron absorbing material can be attributed to its proven performance (over 150 pool years of experience) and the following unique characteristics:
- a. The content and placement of boron carbide provides a very high removtl cross-section for thermal neutrons.
1
- b. Boron carbide, in the form of fine particles, is homogeneously dispersed i throughout the central layer of the Boral panels.
- c. The boron carbide and aluminum materials in Boral do not degrade as a result of long-tenn exposure to radiation.
- d. The neutron absorbing central layer of Boral is clad with permanently bonded l surfaces of aluminum.
- e. Boral is stable, strong, durable, and corrosion resistant.
l Holtec Intemational's QA program ensures that Boral is manufactured by AAR Manufacturing under the control and surveillance of a Quality Assurance / Quality Control Program that conforms to the t : guirements of 10CFR50 Appendix B," Quality Assurance Criteria for Nuclear Power Plants."
As indicated in Tables 3-1 and 3-2, Boral has been licensed by the USNRC for use in numerous PWR and BWR spent fuel storage racks and has been extensively used in miclear installations worldwide.
Boral Material Characteristin Aluminum: Aluminum is a silvery-white, ductile metallic element that is the most abundant in the earth's crust. The 1100 alloy aluminum is used extensively in heat exchangers, pressure and storage tanks, chemical equipment, reflectors and sheet metal i work.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-2 Report HI-981983
I It has high resistance to corrosion in industrial and marine atmospheres. Aluminum has a'n atomic number of 13, an atomic weight of 26.98, a specific gravity of 2.69 and a valence of 3. The physical, mechanical and chemical properties of the 1100 alloy l
aluminum are listed in Tables 3-3 and 3-4.
The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion.
Boron Carbide: The boron carbide contained in Boral is a fine granulated powder that conforms to ASTM C-750-80 nuclear grade Type 111. The material conforms to the chemical composition and properties listed in Table 3-5.
1 References [3.3.2], [3.3.3], and [3.3.4] provide further discussion on the suitability of these materials for use in spent fuel storage module applications.
3.4 Comnatibility with Coolant All materials used in the construction of the Holtec racks have an established history ofin-pool usage. Their physical, chemical and radiological compatibility with the pool environment is an established fact. As noted in Tables 3-1 and 3-2, Boral has been successfully used in both PWR and BWR pools. Austenitic stainless steel (304L) is perhaps the most widely used stainless alloy in nuclear power plants.
i 3.5 Heavy Load Considerations for the Pronosed Reracking Ooeration !
A 100-ton crane will be utilized for handling all heavy loads in the reracking operation. A remotely engageable lift rig, which meets NUREG-0612 stress criteria, will be used to lift the new modules. It consists ofindependently loaded lift rods with a " cam type" lift configuration.
This ensures that failure of one traction rod will not result in uncontrolled lowering of the load, compliant with Section 5.1.6(1) of NUREG-0612. The remotely engageable lift rig also has the following attributes:
SHADED TEXT CONTAINS PROPRIETARY INFORMATION l{oltec International 33 Report 111-981983
l as The stresses in the lift rods are self limiting inasmuch as an increase in the magnitude of the load reduces the eccentricity between the upward force and downward reaction (moment arm).
- b. It is impossible for a traction rod to lose engagement with the lifted rack because j the downward load secures each rod in its locking slot. Moreover, the locked i configuration can be directly verified from above the pool water without the aid of an undenvater camera by the orientation of position locator flags atop each traction rod.
- c. A stress analysis of the rig is performed, the stress limits postulated in ANSI 14.6 (1978) are shown to be met.
- d. The rig is load tested with 300% of the maximum weight to be lifted. The test weight is maintained in the air for one hour. All critical weld joints are liquid penetrant examined, after the load test, to establish the soundness of all critical joints. i l
1 Pursuant to the defense-in-depth approach of NUREG-0612, the following additional measures of safety will be undertaken for the reracking operation.
i'
- a. The cranes and lifting devices used in the project will be given a preventive maintenance checkup and inspection per Oyster Creek plant procedures.
- b. Safe load paths will be developed for moving the new racks in the Reactor l i
Building. The new racks will not be carried over any region of the pool contr.ining fuel.
, c. The rack upending will be carried out in an area which is not poolside.
Additionally, this area will not overlap any safety-related component.
- d. All crew members involved in the rack installation will be given training in the use of the lifting, upending equipment, and all other aspects of rack installation.
- e. Crane stop blocks will be temporarily installed to prevent movement over fuel.
)
SHADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec Intemational 3-4 Report HI-981983
1 In. addition to. the above design, testing, and operation measures, the consequences of a j
postulated rack drop are also considered on the int'egrity of the pool structure. The following analysis is performed.-
l t
- a. The heaviest rack module is postulated to free fall from the top of the water '
. surface level to the pool floor,
- b. The fall of a rack is assumed to occur in its normal vertical orientation, which minimizes the retarding effect of water drag.
- c. The falling rack is assumed to impact the pool slab undergoing an elastic / plastic !
impact.
The results of this calculation show that the postulated rack drop does not cause gross failure of the spent fuel pool slab. Therefore, the integrity of the pool structure is ensured.
The fuel shuffle scheme developed for the spent fuel pool is predicated on the following criteria:
- a. No heavy load (rack or rig) with a potential to drop on a rack shall be carried over or near active fuel. This shall be accomplished by shuffling fuel into racks that are not in the area of the safe load path,
- b. All heavy loads are lifted in such a manner that the C.G. of the lift point is aligned with the C.G. of the load being lifted,
- c. Turnbuckles are utilized to " fine tune" the verticality of the rack being lifted.
All phases of the rack installation will be conducted in accordance with written procedures, which will be reviewed and approved by Oyster Creek.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-5 Report HI-981983 1
p The guidelines contained in NUREG-0612, Section 5 will be followed throughout the rack
' installation. The guidelines of NUREG-0612 call for measures to " provide an adequate defense-in-depth for handling of heavy loads near spent fuel..." and' cite four major causes of load handling accidents, namely:
1
- 1. Operator errors !
( ii. Rigging failure iii. Lack of adequate inspection iv. Inadequate procedures l
l l
The Oyster Creek rack expansion program mitigates the potential for load drop accidents by implementing measures that increase overall quality and safety in the four aforementioned areas. l A summary of the measures specifically planned to deal with the major causes is provided below.
l Operator errors: As mentioned above, Oyster Creek plans to provide comprehensive i
training to the installation crew.
I
) Riggingfailure: The lifting device designed for handling and installation of the racks in l the Oyster Creek fuel pool has redundancies in the lift legs, and lift eyes such that there j are four independent load members. Failure of any one load-bearing member would not '
l lead to uncontrolled lowering of the load. The rig complies with all provisions of ANSI i
14.6 - 1978, including compliance with the primary stress criteria, load testing at 300% of rnaximum lift load, and dye examination of critical welds.
The Oyster Creek rig design is similar to the rigs used in the rerack of numerous other i plants, such as Sequoyah, J. A. Fitzpatrick, Duane Arnold, Three Mile Island Unit 1, D.C. j Cook, and Connecticut Yankee. l
! 4 Lack of adequate inspection: The designer of the racks will develop a set ofinspection I
points which have proven to eliminate any incidence of re-work or erroneous installation in numerous prior rerack projects. Inspection oflifting equipment will be performed per NUREG-0612.
Inadequate procedures: Oyster Creek plans a multitude of procedures to cover the entire rack installation, such as mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance.
The operating procedures planned for the Oyster Creek rack installation are the successors of procedures that were implemented successfully for other projects.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec Intemational 3-6 Report 111981983 !
I In addition to the above, a complete inspection and preventive maintenance program of all the
. cranes and lifting equipment used in the project are planned prior to the start of rack installation.
Safe load paths will be developed as required by NUREG-0612.
Table 3-6 provides a synopsis of the requirements delineated in NUREG-0612 and our intended compliance.
In summary, the measures implemented in the Oyster Creek rack installation are similar to those utilized in all recent reracks in the U.S.
3.6 ' References for Section 3
[3.3.1] " Nuclear Engineering Intemational," July 1997 issue, pp 20-23.
[3.3.2] " Spent Fuel Storage Module Corrosion Repon," Brooks & Perkins Report 554, June 1,1977.
[3.3.3] " Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools," Brooks & Perkins Report 578, July 7,1978.
[3.3.4] "Boral Neutron Absorbing / Shielding Material - Product Performance Report," Brooks & Perkins Repon 624, July 20,1982.
[3.5.1) ANSI N14.6-1978, " Standard for Special Lifting Devices for Shipping Containers Weighing 10000 Pounds or More for Nuclear Materials,"
American National Standard Institute, Inc.,1978.
[3.5.2] ANSI /ASME B30.2, " Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist)," American Society of Mechanical Engineers,1983.
[3.5.3] ANSI /ASME B30.20, "Below-the-Hook Lifling Devices," American Society of Mechanical Engineers,1993.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec International 3-7 Report HI-981983
l Table 3-1 I
BORAL EXPERIENCE LIST - PWR PLANTS i Plant Utility Docket No. Mfg. Year Maine Yankee Maine Yankee Atomic Power 50-309 1977 Donald C. Cook Indiana & Michigan Electric 50-315/316 1979 Sequoyah1,2 Tennessee Valley Authority 50-327/328 1979 Salem 1,2 Public Service Electric & Gas 50-272/311 1980 Zion 1,2 Commonwealth Edison Co. 50-295/304 1980 Bellefonte 1,2 Tennessee Valley Authority 50-438/439 1981 Yankee Rowe Yankee Atomic Power 50-29 1964/1983 Indian Point 3 NY Power Authority 50-286 1987 Byron 1,2 Commonwealth Edison Co. 50-454/455 1988 Braidwood 1,2 Conunonwealth Edison Co. 50-456/457 1988 Yankee Rowe Yankee Atomic Power 10-29 1988 Three Mile Island 1 GPU Nuclear 50-289 1990 Sequoyah (rerack) Tennessee Valley Authority 50-327 1992 Donald C. Cook (rerack) American Electric Power 50-315/316 1992 Beaver Valley Unit 1 Duquesne Light Company 50-334 1993 Fort Calhoun Omaha Public Power District 50-285 1993 Zin 1 & 2 (rerack) Commonwealth Edison Co. 50-295/304 1993 Salem Units 1 & 2 Public Gas and Electric Company 50-272/311 1995 (rcrack)
Haddam Neck Connecticut Yankee Atomic Power 50-213 1996 Company Waterford Unit 3 Entergy Operations, Inc. 50-382 1997 i i
Callaway Union Electric Company 50-483 1997 Gosgen Kernkraftwerk Gosgen-Daniken AG -- 1984 j (Sw;tzerland) l l Koeberg 1,2 ESCOM (South Africa) -- 1985 )
l SH ADED TEXT CONTAINS PROPRIETARY INFORMATION j IIoltec International 3-8 Report HI-981983 i
Table 3-1 BORAL EXPERIENCE LIST - PWR PLANTS Plant Utility Docket No. Mfg. Year Beznau 1,2 Nordostschweizerische Kraftwerke AG -
1985 (Switzerland) 12 Various Plants Electricite de France (France) --
1986 Ulchin Unit 1 Korea Electric Power Company (Korea) -
1995 Ulchin Unit 2 Korea Electric Power Company (Korea) --
1996 Kori 4 Korea Electric Power Company (Korea) --
1996 Yonggwang 1,2 Kc rea Electric Power Company (Korea) -
1996 Sizewell B Nuclear Electric, plc (United Kingdom) --
1997 Angra 1. Furnas Centrais-Electricas SA (Brazil) -
1997 SHADED TEXT CONTAINS PROPRIETARY INFORMATION lloltee International 39 Report 11I 981983 i i
r-Table 3-2 BORAL EXPERIENCE LIST - BWR PLANTS Plant Utility Docket No. Mfg. Year Cooper Nebraska Public Power 50-296 1979 J.A. FitzPatrick NY Power Authority 50-333 1978 Duane Amold towa Electric Light & Power 50-331 1979 Browns Ferry 1,2,3 Tennessee Valley Authority 50-259/260/296 1980 Brunswick 1,2 Carolina Power & Light 50-324/325 1981 Clinton Illinois Power 50-461/462 1981 Dresden 2,3 Commonwealth Edison Company 50-237/249 1981 E.I. Ilatch 1,2 Georgia Power 50-321/366 1981 llope Creek Public Service Electric & Gas 50-354/355 1985 Ilumboldt Bay Pacific Gas & Electric Company 50-133 1985
)
i Lacrosse Dairyland Power 50-409 1976 Limerick 1,2 Philadelphia Electric Company 50-352/353 1980 I Monticello Northem States Power 50-263 1978 Peachbottom 2,3 Philadelphia Electric 50-277/278 1980 Perry 1,2 Cleveland Electric illuminating 50-440/441 1979 Pilgrim Boston Edison Company 50-293 1978 l l
Susquehanna 1,2 Pennsylvania Power & Light 50-387,388 1979 l Vermont Yankee Vermont Yankee Atomic Power 50-271 1978/1986 Hope Creek Public Service Electric & Gas 50-354/355 1989 Shearon liarris Pool B Carolina Power & Light 50-401 1991 Duane Arnold Iowa Electric Light & Power 50-331 1993 Pilgrim Boston Edison Company 50-293 1993 LaSalle 1 Commonwealth Edison Compar y 50-373 1992 Millstone Unit 1 Northeast Utilities 50-245 1989 SilADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec International 3-10 Report Hi-981983
1 1
1 I
i 1
Table 3-2 BORAL EXPERIENCE LIST - BWR PLANTS Plant Utility Docket No. Mfg. Year
~
James A. FitzPatrick NY Power Authority 50-333 1990 Hope Creek Public Service Electric & Gas Company 50-354 1991 Duane Arnold Energy lowa Electric Power Company 50-331 1994 Center Limerick Units 1,2 PECO Energy 50-352/50-353 1994 Shearon Hanis Pool'B' Carolina Power & Light Company 50-401 1996 Nine Mile Point Unit i Niagara Mohawk Power Corporation 50-220 1997 J.A. FitzPatrick NY Power Authority 50-333 1997 (racks added)
Chinshan 1,2 Taiwan Power Company (Taiwan) --
1986 l
Kuosheng 1,2 Taiwan Power Company (Taiwan) --
1991
{
Laguna Verde 1,2 Comision Federal de Electricidad --
1991 (Mexico) l l
SIIADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec International 3-11 Report HI 981983
q l
L Table 3-3 1100 ALLOY ALUMINUM PHYSICAL CHARACTERISTICS Density 0.098 lb/in' 2.713 g/cm' Melting Range 1190'F - 1215*F 643'- 657'C i ThermalConductivity(77 F) 128 BTU /hr/ft 2fpjgg 0.53 cal /sec/cm2 / C/cm Coefficient of Hermal Expansion 4 13.1 x 10 in/in- F (68 F- 212 F) 23.6 x 104 cm/cm *C Specific Ileat (221*F) 0.22 BTU /lb/ F ]'
0.23 cal /g/'C Modulus of Elasticity 10 x 10' psi Tensile Strength (75'F) 13,000 psi (annealed) 18,000 psi (as rolled)
Yield Strength (75'F) 5,000 psi (annealed) j 17,000 psi (as rolled)
Elongation (75 F) 35-45% (annealed) 9-20% (as rolled) liardness (Brinell) 23 (annealed) 32 (as rolled)
Annealing Temperature 650 F 343 C M
SIIADED TEXT CONTAINS PROPRIETARY INFORMATION lloitec International 3-12 Report 11I 981933
r:
. Table 3-4 CHEMICAL COMPOSITION - ALUMINUM (1100 ALLOY) l 99.00% min. Aluminum 1.00% max. Silicone and Iron 0.05-0.20% max. Copper 0.05% max. Manganese 0.10% max. Zinc 0.15% max. Other l
1
(
i SIIADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec International 3-13 Report HI-981983
l Table 3-5 CHEMICAL COMPOSITION AND PHYSICAL PROPERTIES OF BORON CARBIDE CHEMICAL COMPOSITION (WEIGHT PERCENT)
Total boron 70.0 min.
B' isotopic content in natural boron 18.0 Boric oxide 3.0 max.
Iron 2.0 max.
Total boron plus total carbon 94.0 min.
PHYSICAL PROPERTIES Chemical formula B,C Boron content (weight percent) 78.28 %
Carbon content (weight percent) 21.72 %
Crystal structure - rhombohedral Density 0.0907 lb/in' 2.51 g/cm 8 Melting Point 4442*F 2450 C Boiling Point 6332*F 3500'C i
SHADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec Intemational 3-14 Report HI-981983 L. ....... . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
F i
Table 3-6 l HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612)
Criterion Compliance
- 1. Are safe load paths defined for the movement of heavy loads to Yes minimize the potential ofimpact if dropped on irradiated fuel and safe shutdown equipment?
- 2. Will procedures be developed to cover: identification of required Yes equipment, inspection, and acceptance criteria required before movement ofload, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?
- 3. Will crane operators be trained and qualified? Yes
- 4. Will speciallifting' devices meet the guidelines of ANSI 14.6-1978 Yes
[3.5.1] ?
- 5. Will non-customer lifting devices be installed and used in Yes accordance with ANSI B30.20-1971 [3.5.3] ?
- 6. Will the cranes be inspected and tested prior to use in rerack? Yes
- 7. Yes Does the crane meet the intent of ANSI B30.2-1976 [3.5.2] and CMMA-70?
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-15 Report HI-981983
p
! ' 4.0 CRITICALITY SAFETY ANALYSIS l
!' 4.1 Introdetion l l l The high density spent fuel' storage racks for Oyster Creek are designed to assure that the neutron multiplication factor (k er
) is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature correspo'n ding to the highest reactivity. The maximum calculated reactivity includes a margin for I uncertainty in reactivity calculations and in mechanical tolerances, statistically combined, such that the true kar will be equal to or less than 0.95 with a 95% probability at a 95% confidence level. .
Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under credible abnormal conditions, the reactivity will be maintained less than 0.95. The purpose of the present analysis is to confirm acceptability of the rack design.
The fuel selected as the design basis for the racks is a GE 8x8R fuel rod assembly with a maximum (uniform) initial enrichment of 4.6% U-235 containing sufficient gadolinia (Gd2 03 ) to limit the maximum planar k-infinite in the standard cold core geometry to 1.32 or less. Criteria for the acceptable storage of the GE-9B (8x8 rod array with large water rod) and the GE-11 (9x9 rod array with two large water rods) were also developed, based upon the k. in the standard cold core geometry (SCCG)._ The SCCG is defined as an infinite array of fuel assemblies with nmninal dimensions on a 6-inch lattice spacing at 20*C, without any control absorber or voids.
- Applicable codes, standards, and regulations, or peninent sections thereof, include the following:
=
General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.
USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage.
=
USNRC letter of April 14,1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18,1979.
~
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational - 4-1 Report HI-981983 i
USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2
'(proposed), December,1981.
=
ANSI-8.17-1984, Criticality Safety Criteria for. .the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
L. Kopp, " Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants", USNRC Internal Memorandum L. Kopp to Timothy Collins, August 19,1998 To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made:
l The racks were assumed to contain the most reactive fuel authorized to be stored in the facility without any control rods or any uncontained burnable poison, and with the fuel at the burnup conesponding to the highest planar reactivity during its burnup history.
Moderator is pure, unborated water at a temperature within the design basis range l
corresponding to the highest reactivity. 1 l
a Criticality safety analyses are based upon the infinite multiplication factor (k.), i.e., !
lattice of storage racks is assumed infinite in all directions. No credit is taken for i axial or radial neutron leakage, except in the assessment of certain abnonnallaccident conditions where neutron leakage is inherent.
Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4 Report HI-981983
f.
i L
4.2 Summarv and Conclusions l
The fuel assembly selected as the design basis for the racks is a standard GE 8x8R array of BWR fuel rods containing UO2 clad in Zircaloy, and using a uniform initial endchment of 4.6 wt% U-235, evaluated at the maximum reactivity over bumup where the gadolinium is nearly consumed. The i effect of calculational and manufacturing tolerances are added in determining the maximum k.,in the storage rack, and an allowance of 0.01 Ak is also added for possible differences between fuel vendor calculations of the SCCG k. and those reported here.
In BWR fuel, there is a wide variety of designs, including enrichment distribution and gadolinia loading, which often vary in the axial direction. It is conventior.el practice for the fuel vendor,in developing a specific assembly design, to provide values for k., in the. SCCG for each planar (axial) region of significantly different composition or anangements. These k , values are provided at 0%
void (core inlet),40% void (core average), and 70% void (exit condition). The initial Gd2 03 loading enters into the fuel vendor's calculations of the burnup at which the peak reactivity occurs. At this bumup, the gadolinium is essentially depleted. Consequently, calculations of the reactivity in the storage rack do not need to involve gadolinium, but only the average enrichment. Calculations are provided herein illustrating this fact and correlating the k. in the storage rack to the vendor-supplied
- k. in the SCCG. Figure 4.2.1 illustrates the variation in maximum reactivity of the storage rack with values of the k. in the SCCG. The acceptance criteria for safe storage of spent fuel is that the vendor-supplied k. in the SCCG must be equal to or less than 1.32 for the planar region of highest reactivity with fuel of 4.6% average enrichment or less. This cdteria is conservative since (1) all other axial regions (if differences are present) will be lower in reactivity, (2) a planar-average enrichment of 4.6% is higher than would normally be expected in BWR fuel, and (3) the rack calculations use the assembly average enrichment (conservative compared to tne distributed enrichments nonnally used in BWR fuel). Altemative criteria for fuel enrichments less than 4.6%
l are also presented, should these be desired.
l l
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-3 Report HI-981983 i
1 The basic calculations supporting the criticality safety of the Oyster Creek fuel storage racks for the s
I design basis fuel are summarized in Table 4.2.1. For the design basis fuel,4.6% enrichment or less with a k. less than or equal to 1.32 in the SCCG, the storage racks conform to the USNRC criterion I
(
of a maximum k,,less than or equal to 0.95. l Calculations were also made for fuel of GE-9B (8x8) and GE-11 (9x9) designs in the storage rack l
(
and in the SCCG,' confirming that the maximum k of 1.32 in the standard cold core geometry is l acceptable for fuel of various design concepts. Since the limiting k. in the SCCG is enrichment dependent, calculations were also made for lower-enriched fuel. Results are summarized in Figure j 4.2.1, illustrating that a SCCG k. of 1.32 is acceptable for all enrichments (up to an average of 1
4.6%). At enrichments less than 4.6%, a higher SCCG k would be acceptable as illustrated m
{
Figure 4.2.2. For enrichments less than 3.2%, the maximum k,y in the rack is less than 0.95 (including uncertainties and allowances) regardless of gadolinia content and the k. in the SCCG need not be considered.
The criteria for spent fuel to be acceptable for storage in the Oyster Creek spent fuel racks are the following:
Any fuel assembly which has a planar-average enrichment of 3.2% or less, or Fuel assemblies with a planar SCCG k of 1.32 or less, with an average enrichment of 4.6% or less, or Alternatively, any fuel whose enrichment-SCCG k. combination falls within the acceptable domain (below the curve) of Figure 4.2.2.
i SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-4 Report HI-981983
m 4.3 Abnormal and Accident Conditions None of the abnormal or accident conditions that have been identified as credible will result in exceeding the limiting reactivity (km of 0.95). The effects on reactivity of credible abnormal and accident conditions are summarized in Table 4.3.1. The double contingency principle of ANSI N16.1-1975 (and the USNRC letter of April 1978) specifies that it shall require at least two unlikely independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions. Other hypothetical events were considered and no credible occurrences or configurations have been identified that might have any adverse effect on the storage rack criticality safety.
l t
i l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-5 Report HI-981983 i
i i
l
p 4.4 Innut Parameters 4.4.1 Design Fuel Assembly Soecifications The design basis fuel assembly is a standard GE 8x8R array of BWR fuel rods containing UO2 clad in Zircaloy (62 fuel rods with 2 water rods). For the nominal design case, fuel of uniform 4.6 wt%
U-235 enrichment was taken as the principal design criterion. The GE-9B design, an 8x8 array with a single large water rod, and the GE-11 fuel design, a 9x9 array of fuel rods with 7 rods replaced by two Zircaloy water rods, were also evaluated. Design parameters for all three types of fuel are summarized in Table 4.4.1. Fuel specification of the early 7x7 designs were also evaluated and found to be less reactive than the reference design fuel (primarily because of the lower enrichments employed).
4.4.2 Storage Rack Cell Specifications The design basis storage rack cell consists of an egg-crate structure, illustrated in Figure 4.4.1, with fixed neutron absorber material (Boral) of 0.0162 g/cm2 boron-10 areal density ( ; gms B-10/cm 2 minimum) positioned between the feel assembly storage cells in a inch channel. This arrangement provides a nominal center-to-center lattice spacing of 6.106 inches. Manufacturing tolerances, used in evaluating uncertainties in reactivity, are indicated in Figure 4.4.1. The 0.075-in stainless-steel box which defines the fuel assembly storage cell, has a nominal inside dimension of 5.93 in. This allows adequate clearance for inserting / removing the fuel assemblies, with or without the Zircaloy flow channel.
l l
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-6 Report HI-981983
p !
l i
l l
l 4.5 Analvtical Methcxiolog_v 4.5.1 Computer Codes and Benchmarking In the fuel rack evaluation, criticality analyses of the high density spent fuel storage racks were performed with the CASMO4 code [4.5.1], a two-dimensional multi-group transport theory code.
Independent verification calculations were made with the MCNP code [4.5.2], a continuous energy Monte Carlo code developed by the Los Alamos National Laboratory, and with the KENO-Sa code package [4.5.3] using the 238-group SCALE' cmss-section library with the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment).
l Benchmark calculations are presented in Appendix 4A and indicate a bias of 0.0009
- 0.001i for MCNP and 0.003010.0012 (95%/95%) for NITAWL-KENO-5a. In the geometric model used in the calculations, each fuel rod and its cladding were explicitly described and reflecting boundary conditions (zero neutron current) were used in the axial direction and at the equivalent centerline of the Boral and steel plate between storage cells. These boundary conditions have the :ffect of creating an infinite array of storage cells in all directions.
The CASMO4 computer code was used as the primary method of analysis as well as a means of evaluating small reactivity increments associated with manufacturing tolerances. Bumup calculations '
were also perfonned with CASMO4, using the restart option to describe spent fuel in the storage l cell. MCNP and KENO-Sa were used to assess the reactivity consequences of eccentric fuel positioning, channel bulging, and abnormal locations of fuel assemblies. ;
I I
l SCALE is an acronym for Standardized Computer Analysis for Licensing Evaluation, a standard cross-section set developed by the Oak Ridge National Laboratory for the USNRC.
[
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-7 Report HI-981983 l
1 l
t 4.5.2 CASMO4 Validation
' The CASMO4 calculations were validated (benchmarked) against MCNP and KEN 05a for the specific fuel and geometries involved. Comparison calculations are listed in Table 4.5.1 for fuel of at 3.1% enrichment. (MCNP and KENO 5a results are bias corrected and listed at the 95%/95%
level)'
1 l
These data confirm the validity of the CASMO4 calculations. In addition, the data indicate that, for 3.1% enrichment, without any depletion, the three fuel types exhibit very nearly the same reactivity.
Other calculations, described later, show that at a k. of 1.32 in the SCCG (burned fuel), the GE-98 and GE-1I show slightly higher reactivities.
j 4.5.3 Gadolinia Effects and Burnup i
Gadolinia is used in all BWR fuel designs (other than some early concepts with relativity low I enrichments) as a means of augmenting reactivity control in core operations. Gadolinium has a higher cross section than U-235, and thus, the reactivity of an assembly increases with burnup, reaching a maximum at some point in burnup where the gadolinium is virtually depleted. For fuel of 4.6% average enrichment, Figure 4.5.1 illustrates the reactivity variation with burnup for several illustrative gadolinia loadings with 8x8R fuel of 4.6% enrichment, evaluated in the spent fuel storage racks (bias and uncertainties included). Three of the example fuel assemblies shown in Figure 4.5.1 would be acceptable for storage and one would not (the 6Gd 1% E 4.6%, presented for illustrative purposes only).
Associated with each of the fuel assemblies in Figure 4.5,1 are values of the k. in the standard cold core geometry, Figure 4.5.2 is a correlation of the k. in the SCCG and the corresponding k.in the storage rack for 8x8R fuel of 4.6% enrichment. With this type of correlation, the variation in rack K-factor for one-sided tolerance at 95%/95% from NBS Handbook 91 [4.5.4] is 1.71.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec Intemational 4-8 Report HI-981983
1 i
l i
reactivity is effectively independent of the gadolinia loading. Also shown in Figum 4.5.2 is the correlation curve calculated without any gadolinia. This latter curve bounds the cases calculated with gadolinia and illustrates that assemblies with a maximum k. of 1.32 in the SCCG will be acceptable I for storage in the racks. Figure 4.5.3 presents the same curve (calculated without gadolinia) with maximum k., points from Figum 4.5.1 superimposed. Any other gadolinia loading would fall on the same curve, at a point corresponding to the [ vendor supplied] k. in the SCCG.
Similar calculations without gadolinia were made for fuel of other enrichments and the results am summarized in Figure 4.2.1 and also include the GE-9B and GE-1I fuel types. For the same value i of k. in the SCCG, the GE-9B and GE-11 fuel types show a slightly higher reactivity in the rack than the 8x8R fuel. Similarly, fuel of enrichments less than 4.6% show lower values of k. in the rack for :he same k,. in the SCCG. This permits a higher limiting k. in the SCCG for lower enrichments and this fact is reflected in Figure 41 !
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-9 Report HI-981983
e 4.6 Criticality Analyses and Tolerance Variatiana 4.6.1 Nominal Desian Case The reactivity of fuel assemblies with enrichments ranging from 3.6% to 4.6% were pmviously shown in Figure 4.2.1. At a k. (SCCG) of 1.32, the infinite multiplication factor, k , in the storage racks is given in Table 4.2.1, including all known uncenainties and all allowances, confirming that the maximum k. is less than the regulatory limit (k,y< 0.95). Reactivity effects of natural uranium blankets, normally located at the ends of the assemblies, wem { conservatively] neglected since an infinite fuel length corresponding to the most reactive planar segment was used.
Calculations were also made at lower enrichments at zero burnup to determine the highest enrichment that would be acceptable for storage without consideration of bumup, gadolinia or the SCCG k.. These results, with all uncertainties and allowances included, are listed below, confimiing that any fuel with an average enrichment of 3.2% or less is acceptable for storage without credit for burnup or gadolinia that would normally be present. This enrichment criteria is also applicable to GE-9B and GE-11 type fuel assemblies. The total correction for tolerances and uncertainties is (0.0073 for toleroces + 0.0033 for channel bulging) = 0.0106 akt -
3.2%E Ref. 8x8R Fuel = 0.9373 @ 3.2%E with 0.0106 ak correction GE-9B 8x8 Fuel = 0.9348 @ 3.2%E with 0.0106 ak correction GE-119x9 Fuel = 0.9357 @ 3.2%E with 0.0106 ak correction GE-119x9 Fuel (top) = 0.9365 @ 3.2%E with 0.0106 ak correction (w/ missing rods)
For direct calculations without reference to vendor calculations, the 0.01ak allowance is not applicable.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-10 Repon HI-981983 I
u
\
i 1
GE-11 fuel has part-length fuel rods. In the region near the top of the fuel, the reactivity is slightly )
higher than the fully-rodded region below. The reactivity effect of the part-length fuel rods is also included in the final evaluation (Table 4.2.1). If axial leakage were to be included, the reactivity would be lower. These data confirm that any fuel with an enrichment of 3.2% or less is acceptable i
l for storage in the racks, regardless of the k in the SCCG, the fuel burnup or fuel rod array.
4.6.2 Uncedainties Due to Manufacturing Tolerances The reactivity effects associated with manufacturing tolerances are listed in Table 4.6.1 and discussed below. These tolerances were evaluated at i1 MWD /KgU burnup, which corresponds approximately to a k., in the SCCG of 1.33. The enrichment tolerance was conservatively evaluated at 3.2%.
4.6.2.1 Boron Loading Variation The Boral absorber panels used in the storage cells are nominally 0.070-inch thick, with a B-10 areal 1
density of 0.0162 g/cm2 . The manufucturing tolerance limit is -
g/cr5 in B-10 content, including both thickness and composition tolerances. This assures that the minimum boron-10 areal density will not be less than g/cm2 . Differential CASMO4 calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of 10.0032 ak.
4.6.2.2 Boral Width Tolerance Variation The reference storage cell design uses a Boral panel width of 5.00 inches. The tolerance on the Boral !
l width is inch. Calculations (CASMO4) showed that this tolerance corresponds to a 0.0017 ak uncertainty.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec Intemational 4-11 Report HI-981983
F 4.6.2.3 Storage Cell I*tice Pitch Variation The design storage cell lattice spacing between fuel assemblics is 6.1% in. Decreasing the lattice pitch increases the reactivity. For the manufacturing tolerance of ; -inches, the corresponding uncenainty in reactivity is t 0.0013 ak as determined by differential CASMO calculations.
l 4.6.2.4 Stainless Steel Thickness Tolerances i
The nominal thickness of the stainless steel box is 0.075 inches and 0.0235 inches for the steel sheath.The maximum positive reactivity effect of the expected stainless steel thickness tolerances j was calculated (CASMO4) to be 10.0016 ak.
4.6.2.5- Fuel Enrichment and Density Variation CASMO4 calculations of the sensitivity to small enrichment variations yielded an average coefficient of 0.0040 ak for the enrichment tolerance of 0.05 wt% U-235, in the enrichment range from 3.0 to 3.2% enrichment. For the estimated tolerance in percent U-235 enrichment, the maximum uncenainty is t 0.0040 Ak; this becomes smaller for higher-enriched fuel and is not significant at enrichments below 3.0% (because the considerably lower reactivity).
Calculations were also made to determine the sensitivity to the tolerance in UO 2fuel density (--
g/cc). These calculations indicate that the storage rack k is increased by 0.0021 Ak for the expected maximum 10.61 gms/cc stack density. Lower fuel densities result in correspondingly lower values of reactivity. Thus, the maximum uncertainty due to the tolerances on UO 2 density is i 0.0021 ak.
4.6.2.6 Zirconium Flow Channel Elimination of the zirconium flow channel results in a small decmase in reactivity. More significant F is a positive reactivity effect resulting from potential bulging of the zirconium channel, which moves SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec international 4-12 Report HI-981983 l
l l
(
f- 1 i
i l
i the channel wall outward replacing water near the Boral absorber. For the maximum expected l
bulging based on estimates provided by GE and conservatively assumed to be uniform throughout j all assemblies, an incremental reactivity of + 0.0033 Ak could result (determined by differential l i
MCNP calculations). Although the bulging penalty would not be expected to occur uniformly everywhere, the maximum penalty was conservatively tmated as an additive factor, rather than being statistically combined with other uncertainties.
4.6.3 Uncertainty in Deoletion Calculations Since critical experiment data with spent fuel is nN available for detennining the uncertainty in depletion calculations, an allowance for uncertainty in reactivity was assigned based upon other 1
considerations. In the Oyster Creek racks, the reactivity decrement in the absence of gadolinium is approximately 0.08 ak. Assuming the uncertainty in depletion calculations is less than 5% of the l
total reactivity decrement, an uncertainty in reactivityt equal to 0.0040 ak would result.
1 Only that ponion of the uncertainty due to depletion calculations. Other uncertainties are accounted for elsewhere.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Iloltec Intemational 4-13 Report 14981983 I
b 4.7, AtsumL0nd Accident Conditions
-4.7.1 ; Igngaggipre and Water Density Effects
. The moderator temperature coefficient of reactivity is negative. Using the minimum temperature of 4*C (maximum possible water density), therefore assures that the true reactivity will always be lower than the calculated value mgardless of the temperature. Temperature effects on reactivity have been calculated and the results are shown in Table 4.7.1. Introducing voids in the water in the storage cells (to simulate boiling) decreased reactivity, as shown in the tabic. Boiling at the submerged depth of the racks would occur at approximately 120*C.
4.7.2 Abnormal Location of a Fuel Assembly It is hypothetically possible to suspend a fuel assembly of tlie highest allowable reactivity outside and adjacent to the fuel rack, although such an accident condition is highly unlikely. The exterior walls of the rack modules facing the outside (where such an accident condition might be conceivable) is a region of high neutron leakage. The k,y with an extraneous fuel assembly of the maximum reactivity, located outside and adjacent to the fuel rack, is essentially the same as the k,y without the extraneous fuci assembly present. With neutron leakage included, this k,n is less than the k. used as the reference. Thus it is concluded that the abnonnal location of a fuel assembly will have a negligible reactivity effect.
4.7.3 Eccentric Fuel Assembiv Positionin_e The fuel assembly is normally located in the center of the storage rack cell with bottom fittings and 9 spacers that mechanically restrict lateral movement of the fuel assemblies. Nevertheless, calculations with the fuel assembly moved into the comer of the storage rack cell (four-assembly cluster at closest approach), resulted in a small negative reactivity effect. Thus, the nominal case, with the fuel assembly positioned in the center of the storage cell, yields the higher reactivity.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-14 Report HI-981983
l l
l
, i
, 14.7.4, ,
Zigonium Fuel Channel Distortion L'
Consequences of bulging of the zirconium fuel channel are treated as a mechanical deviation in Section 4.6.2.6. Bowing of the zirconium channel (including fuel rods) results in a local negative mactivity effect analogous to that of eccentric positioning the fuel assembly toward one side of the.
Thus, any bowing that might occur will result in a reduction in mactivity. ;
4.7.5 Dropoed Fuel Assembly i
For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of the rack with 4 a minimum separation distance from the active fuel mgion of more than the 12 inches, which is sufficient to pmclude neutron coupling (i.e., an effectively infinite separation). Maximum expected deformation under seismic or accident conditions will not reduce the minimum spacing to less than 12 inches.
It is conceivable that a dropped assembly might penetrate a storage cell in a vertical position, impacting and causing local deformation of the base plate. Analysis (Chapter 7) indicates that the maximum local deformation is 2.26 inches, with smaller deformations in the immediately adjacent cells. Conservative calculations, using a bounding deviation of 2.5 inches of exposed fuelin all cells everywhere, showed that the k-eff of the rack was less than the reference k-infinite and i therefore there would be no increase in the design basis reactivity, Consequently, fuel assembly drop accidents will not result in a significant increase in reactivity.
4.7.6 Fuel Rack Imaral Movement
)
.Normally, the individual rack modules in the spent fuel pool are separated by a water gap which would normally eliminate the need for poison panels between rack modules. However, as an added i precaution against possible mis-alignment or seismically-induced movement, Boral panels are L
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-15 Report HI-981983 1
r i
1 installed in the rack wall along both sides of the water gap. With this configuration, the maximum reactivity of the storage rack is not dependent upon the water-gap spacing between modules. Thus, misalignment of the racks or seismically induced movement will not affect the reactivity of the rack.
I 1
l 1
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 4-16 Report HI-981983
f l 4.8 References for Section 4 i
) [4.5.1] A. Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly Burnup l Program," AE-RF-76-4158, Studsvik report (proprietary).
A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion 4
Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604,1977.
D. Knott, "CASMO-4 Benchmark Against Critical Experiments", . Studsvik i Repon SOA-94/13 (Proprietary). ;
M. Edenius, et. al., "CASMO-4 A Fuel Assembly Burnup Program User's 1 Manual" Studsvik Repon SOA/95/1.
[4.5.2) J.F Briesmeister, Ed., "MCNP- A General Monte Carlo N-Panicle Transpon Code, Version 4A",los Alamos National Laboratory, LA-12625-M (1993).
[4.5.3] N.M. Greene, et.,al., "NITAWL-II: Scale system Module for Performing Resonance Shielding and Working Library Production", in SCAIF: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation. NUREG/CR-0200, Rev. 2, December 1984, L..M. Petrie and N.F. Landers, " KENO 5a - An improved Monte Carlo Criticality Program with Supergrouping:, in SCAIF: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation. NUREG/CR-0200, Rev. 2 December 1984.
[4.5.4] M.G. Natrella, Exocrimental Statistics. National Bureau of Standards, Handbook 91, August 1963.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational- 4-17 Report HI-981983
..a
i- J Table 4.2.1 l
l
SUMMARY
OF CRITICALITY SAFETY ANALYSES Temperature assumed for analysis 4*C Fuel Enrichment (average) 4.6
- k. in SCCG 1.32 Reference CASMO k.,
8x8R Fuel 0.9184 GE-98 Fuel 0.9202 GE-11 Fuel 0.9206 GE-11 Fuel (top, w/ missing rods) 0.9232 Uncertaintin Removal of flow channel (see Section 4.6.2) negative Eccentric assembly location (see Section 4.7.3) negative Tolerances (Table 4.6.1 and Section 4.6.2) 0.0061 Uncertainty in Depletion calculations (see Section 4.6.3) 0.0040 Statistical Combination' O.0073 Effect of Channel Bulge (see Section 4.6.2) + 0.0033 Allowance for Vendor Calculations + 0.0100 l
Maximum reactivity 8x8R 0.9317 i 0.0073= 0.9390 GE-9B Fuel 0.9335 0.0073= 0.9408 ;
GE-11 Fuel 0.9339 0.0073= 0.9412 l GE-11 Fuel (top) 0.936510.0073= 0.9438 i
Square root of sum of squares of all independent tolerance effects.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION fioltec International 4-18 Report HI-981983 l
Table 4.3.1 REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS 1
Accident / Abnormal Condition Reactivity Effect Temperature increase Negative (Table 4.7.1)
Void (boiling) Negative (Table 4.7.1) l Assembly dropped on top of rack Negligible Misplacement of a fuel assembly - Negligible l Seismic Movement Negligible SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-19 Report HI-981983
1 1
l Table 4.4.1 FUEL ASSEMBLY DESIGN SPECIFICATIONS FUEL ROD DATA 8x8R GE-9B GE-11 Cladding outside diameter, in. 0.483 0.483 0.440 Cladding inside diameter,in. 0.419 0.419 0.384 !
l Cladding material Zr-2 Zr 2 Zr-2 Pellet diameter, in. 0.410 0.411 0.376 Enrichment (design basis) 4.6 0.05 4.6 0.05 4.6 0.05 UO 2density (stack), g/cc UO2 10.41 2 0.20 10.41 0.20 10.41 0.20 WATER ROD DATA I
l Number of Water Rods 2 1 2 l Inside diameter, inch 0.414 1.260 0.920 Outside diameter 0.484 1.340 0.980 1 Material Zr-2 Zr-2 Zr-2 :
FUEL ASSEMBLY DATA Fuel rod array 8x8 8x8 9x9 Number of fuel rods 62 60 74 Fuel rod pitch, inch 0.640 0.640 0.566 l Fuel channel, material Zr-2 Zr-2 Zr-2 '
Inside dimension, inch 5.278 5.278' 5.278' Outside dimension, inch 5.478 5.438 5.438 Actual fuel channel is 63 mil thick with 100 mil on coi.v.rs. Ccaservative average of 80 mil used.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-20 Report HI-981983
Table 4.5.1 COMPARISON OF CALCULATIONAL METHODS
@ 3.1% ENRICHMENT CASMO KENO 5a MCNP GE 8X8R fuel Calculated k. 0.9188 0.9131 i 0.0003 0.9134 2 0.0003 Corrected' k. -
0.9161 0.0013 0.9143 0.0012 GE-9B (8x8) l Calculated k. 0.9164 0.9121 0.0003 0.9138 0.0003 Corrected' k. -
0.9151 0.0013 0.9147 0.0012 GE-1I (9x9)
Calculated k. 0.9172 0.9139 0.0003 0.9152 0.0004 Corrected' k. -
0.9169 0.0013 0.9161 0.0013 l
l i
l Includes Monte Carlo statistics and uncertainty in bias, at 95% probability,95% confidence level (one- sided),
)
where the k-factor is 1.71[4.4].
j SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-21 Repon HI-981983 i
)
l Table 4.6.1 REACTIVITY UNCERTAINTIES DUE TO MANUFACTURING TOLERANCES i Nominal Quantity Value Tolerance k., l Boron Loading 0.0162 g/cm 2 g/cm 2 10.0032 1
Boral width 5.00 inches inches 10.0017 Lattice spacing 6.106 inches inches i0.0013 SS thickness 0.075 /0.0235 inches 0.0016 Fuel enrichment @ 3.17s U-235 10.05% U-235 0.0040' l
3 l Fuel density 10.41 g/cm g/cm 3 10.0021 l
Statistical combination" 0.0061 of tolerance uncertainties l
l 1
Effect of enrichment tolerance is significantly smaller at higher enrichments.
Square root of sum of squares of the independent tolerance effects.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec Intemational 4-22 Repon HI 981983 ,
B I
l Table 4.7.1
)
EFFECT OF TEMPERATURE AND VOID ON CALCULATED '
REACTIVITY OF STORAGE RACK I l
1 Case Incremental Reactivity Change, ak !
i 4*C (39"F) Reference !
15.5"C (60*F) -0.0013 20*C (68 *F) -0.0025 50*C (122*F) -0.0074 110 C (230*F) -0.0178 120 C (248 F) -0.0233 120*C + 10% void -0.0452 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-23 Report HI-981983
0.96 GE- 9B Fuel 4. 5 si; (Solld) ' N c' fI l GE--11 Fue' (top) (.6 EE
l
- (Doshod) N %, /
0.94
,/
/
8xB4 6mE / -8x8 Fl Jet 6.1 sE
$ -0.93 \ '# ! #
)[ / /
-- U.92 m _ ,' (f f f e . /
ACCEPTABLEDOMAIN 5 -
(Below Curve) ,
.5 '
0.91 y/ A i :
~
i f 8xt' 3. 5 sl: j
.x 0.90 f f
/ f
. /
/
,/ j 0.89 /
.j i 0.88 / /
- /
1
/
~
0.87 ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,,
1.24 1.26 1.28 1.30 1.32 1.34 1.36 1.38 1.40 k-Infinite in the Standard Cold Core Geometry Fig. 4.2.1 Correlation of the k-infinite in the Standard Cold Core Geometry with the k-infinite in the Storage Rock
r:
~
l l
4 1.365 1.360 \% i 0
+e 1.353 k O ep e 1.350 \
- og ,
i
!c N&
- 8
- 1.345
- vo,r w 1.340 - A .
e : ( of E - Ye O
- 3 i 0 -
E 1.335 (A I
.s o, ;
Y"9 \ l E 1.330 2 _ N N
- Acc eptab le Do main N 1.325 s 1.320 ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,,
3.20 3.40 3.60 3.80 4.00 4.20 4.40 4.60 Initial Enrichment, wt z U-235 1
Fig. 4.2.2 Limiting k-infinite in the Standard Cold Core Geometry for Fuel Enrichments between 3.25 and 4.6z
I I
I l
l l
l l
l BORAL i
! 0.0162 g B-10/cm 2 l
CENTERLINE THRU 5.00 2 1/16" WIDE CENTER OF BORAL ~ 0.070
- 0.004" Thk ~ +- SS CELL l 4 SIDES IN 0.077" CHANNEL 5.9505
- 0.03" ID l C
~~~~~~~~-~~~~~---~~ [ 0.075"/0.0235" Th 0 04ns >
- ,..i ?
I '
Example: 9 x 9 Square Array $
- l0006
~
Pellet OD 0.376" -
fuel Density 957. T.D. !
Clad ID 0.384" . E Clod 00 0.440" .
t 0.08775" -
. s 1
! ! i OOO 's '
Chan.,0 S m e-Chan. OD 5.437
- l E
.: g E
~
157 0000006. ! !
s 0000000h : i s
2 0 0 0 0 0 0 0 0 t', ! !
l 4 . . . .nx ,nx . . o,n,n. , ,n, <nn, .,,,nx ~,,
t j mwww swwwxw j NOT TO SCALE I .I
. .I I. l j
il 5.9305
- 0.03" CELL ID ;! ! j iI ;
6.106
- 0.03" LATTICE SPACING 1 'l l
l l
l C C l 6 I
l l
Fig. 4.4.1 CROSS-SECTION OF TYPICAL STORAGE CELL !
(CALCULATIONAL MODEL) l l
[1- ..
I l
0.97 0.96 0.95 8 S l
- g ^ !
0.94 ~ l a
c
- l 5 0.93: e"b g y*
T -
- 0.92 3
- 0. 91 ' C if #
tb j 5
)- Ob
- j 0.90 - , t c
$ 0.89 y 3
l 0.88 f M h i
" : h
!e 0.87 -
s g .
j 0. 86 _" g-
- b 0.85 {
0.84 !
0.83 ~ , ,,, ,,,, , , , , , , , ,
0 5 10 15 20 Fuel Burnup, MWD /Kgu i
l Fig. 4.5.1 Variation in the Rock Reactivity for Several lliustrative Examples of Gd 203 Loadings
1.04 1.02 . -
- / i NO G400LINIA ,
1.00 '
, )
,', l 0.98 ' i e : ,
e E 0.96
- g' h : /
i 8
ac 0.94
^ /
e a' : E Gd' sE4.6s m
$ 0.92
$t c -
2- 0.90
/ '6 Gd3s E4.6s t; .
8
- 00 - )
Nil Gd 22 E4.6s
~
0.86 - ,'
E Gdfim E4.6s 0.8A_, , ' , g, 0 . 82 ' , , , , , , , , , , , , , , , , ,,,, ,,,, ,,., ,,,, ,,,, ,,,, ,,,,
1.18 1.20 1.23 1.25 1.28 1.30 1.33 1.35 1.38 1.40 1.43 1.45 k-infinite in the Standard Cold Core Geometry I
i l Fig. 4.5.2 Correlation of k-infinite in the Standard Cold l l
i Core Geometry and k-infinite in the Storage Rock I
_a
1.03_
1.02: 7 i /
1.01 1.00 >
/
^ -
S -
0.99
_=
c : ,
e 0.98f o
- j
/
g 0.97_ o on mn mu,oo , to to 5 :
$ 0.96 o :
if 0.95 -
e : /
- / 8 Roi e of 2s Geoo t tn to
& 0.94 1 i /
0.93 ~ '
I, b 6 Rods of Js Gedol in te
.c 0.92: 7 E i /
s 0.91_
? :
/
0.90 0.89 ~
i 0 Rod,i of Es God >l in to 1 0.88 j
~
]
1 0.87 ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,iiiii.... , , , , , , , , , .ii.., , , , , , , , , , , , , , , , , , , ,,,,,,,,,
i 1.23 1.25 1.27 1.29 1.31 1.33 1.35 1.37 1.39 1.41 1.43 1.45
]
k-Infinite in the Standard Cold Core Geometry Fig. 4.5.3 Limiting Correlation of k-infinite in the SCCG and k-Infinite in the Storage Rock (with examples of Gd203 Leadings Superimposed)
I l
i
r-l
, APPENDlX 4A: BENCHMARK CALCULATIONS j 4A.1 INTRODUCTION AND
SUMMARY
Benchmark calculations have been made on selected critical experiments, chosen, in so far i as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A.1] is a continuous energy Monte Carlo code and KEN 05a [4A.2]
uses group-dependent cross sections. For the KENO 5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II [4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errorst (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.
In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the "B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.
Table 4A.1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in
{
subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain ;
criticality; some other parameter or parameters must be concurrently varied to maintain criticality. ;
One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KEN 05a computes and prints the " energy of the average iethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KEN 05a, the number of fissions in each group may be collected and the EALF determined (post-processing).
Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectmm when the spectmm should be different for the
' various cases analyzed, as evidenced by the spectrum indices.
Holtec Report HI-981983 Appendix 4A, Page 1
n Figures 4A.1 and 4A.2 show the calculated k,, for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO 5a, respectively (UO2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental errort in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the I PNL criticals.
Linear regression analysis of the data in Figures 4A.1 and 4A.2 show that there are no I trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KENO 5a). The total bias (systematic error, or mean of the deviation from a '
k,y of exactly 1.000) for the two methods of analysis are shown in the table below.
Calculational Bias of MCNP4a and KENO 5a MCNP4a 0.0009 i 0.0011 KENO 5a 0.0030 i 0.0012 l s
The bias and standard error of the bias were derived directly from the calculated k,y values in Table 4A.1 using the following equations", with the standard error multiplied by the one-sided K-factor for 95% probability at the 95% confidence level from NBS Handbook 91 [4A.18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2),
f= k, (4A.1)
' A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.
These equations may be found in any standard text on statistics, for example, teference
[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO 5a.
Holtec Report HI-981983 Appendix 4A, Page 2
l l
n n
[ k,2 - ([ k,)2 jn j g 2, 4i s.: (4A.2) n (n-1)
Bias = (1- E ) Ko r (4A.3) where ki are the calculated reactivities of n critical experiments; og is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias !
(mean)); K is the one-sided multiplier for 95 % probability at the 95 % confidence level (NBS Handbook 91 [4A.18]).
Formula 4.A 3 is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( 1- E ), is the actual bias which is added to the MCNP4a and KENO 5a results.
The second :erm, Kog, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95% probability at the 95% confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO 5a are 2.04 and 2.05, respectively.
The bias values are used to evaluate the maximum km values for the rack designs, i KENO 5a has a slightly larger systematic error than MCNP4a, but both result in greater precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section sets in KENO 5a (SCALE) calculations.
4A.2 Effect of Fnrichment n
The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and 4A.4 show the calculated k, values (Table 4A.1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms l that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO 5a). Thus, there are no corrections to the bias for the various enrichments, i-I Holtec Report HI-981983 Appendix 4A, Page 3
~
l As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENO 5a for various enrichments.
The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg.' Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of k, for the two independent codes as evidenced by the 45* slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.
4A.3 Effect of OB Londing Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment), i the reactivity worth of the absorbers in the PNL tests is very low and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.
i Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A.1) and shows the reactivity worth (Ak) of the absorber.t ,
1 No trends with reactivity worth of the absorber are evident, although based on the ;
calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.
To further confirm the absence of a significant trend with B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO 5a (as suggested in Reg.
Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.
These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45' line, within an expected 95 % probability limit).
The reactivity wonh of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.
Holtec Report HI-981983 Appendix 4A, Page 4
o l
1
' 4A.4 ,
Miecallananus and Minor Parameters 4A.4.1 Reflector Matarial and Soncinn PNL has performed a number of critical experiments with thick steel and lead reflectors.t ,
Analysis of these critical experiments are listed in Table 4A.5 (subset of data in Table 4A.1). There appears to be a small tendency toward overprediction of k, at the lower spacing, although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.
4A 4.2 Fuel Pellet Dinmeter and Lattice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, j the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.4% to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel. Thts, the critical experiments analyzed provide a reasonable representation of power reactor fuei. Based on the data in Table 4A.1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least ,
over the range of the critical experiments applicable to rack designs.
]
4A 4.3 Soluble Boron concentration Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO 5a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher solub!e boron concentrations could be slightly conservative.
Paraliel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not peninent to the Holtec rack design.
Holtec Report 10-981983 Appendix 4A, Page 5
4A.5 MOX Fuel l The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for UO2fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a y of 1.00, indicating that when Pu is present, both MCNP4a and KENO 5a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, l the KENO 5a calculated reactivities are below 1.00, suggesting that a small trend may exist
- with KENO 5a. It is also possible that the overprediction in Q for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This l
l possibility is supported by the consistency in calculated 4 over a wide range of the
- spectral index (energy of the average lethargy causing fission).
l l
l l
1 l
l l
Eltec Report HI-981983 Appendix 4A, Page 6 i
L
i 4A.6 References
[4A.1] J.F. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).
[4A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.
I l
[4A.3] M.D. DeHart and S.M. Bowman, " Validation of the SCALE Broad i Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in !
Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460) l Oak Ridge National Laboratory, September 1994. '
[4A.4] W.C. Jordan et al., " Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Systems, Inc., Oak Ridge National l
Laboratory, December 1986. '
[4A.5] O.W. Hermann et al., " Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated,
[4A.6] R.J. Larsen and M.L.. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall,1986.
[4A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.
[4A.8) G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW-1645-4, Babcock & Wilcox Company, November 1991.
[4A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984.
Holtec Report HI-981983 Appendix 4A Page 7
f I 1
[4A.10] J.C. Manaranche et al., " Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods," Trans.
j Am. Nucl. Soc. 33: 362-364 (1979).
l [4A.11] S.R. Bierman and E.D. Clayton, Criticality Experiments with l Suberitical Clusters of 2.35 w/o and 4.31 w/o 235U Enriched UO2 l Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.
[4A.12] S.R. Bierman et al., Criticality Experiments with Suberitical Clusters of 2.35 w/o and 4.31 w/o 225 U Enriched UO2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December,1981.
[4A.13] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o 235 U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.
[4A.14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest !
Laboratory, April 1990. I
[4A.15] B.M. Durst et al., Critical Experiments with 4.31 wt % 235U l Enriched UO2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.
[4A.16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.
[4A.17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1%5.
[4A.18] M.G. Natrella, Exnerimental Statistics, National Bureau of Standards, Handbook 91, August 1%3.
I Holtec Report HI-981983 Appendix 4A, Page 8 i
a 9 5 3 6 9 6 9 7 2 5 6 4 3 1 8 e O 7 5 4 4
3 9
2 4
9 4 9 4 6 6 8 1 1 2 0 g
a
) N 1 2 1 1 1 1 6
1 9
1 9 0 7 0 7 P
Ve E 0 0 0 0 0 0 0 0 1 2 1 2 1
( K 0 0 0 0 0 A
' 4 S
I a i x
4 9 3 9 d A T 5 5 9 2
2 3
1 1
3 8
1 8
8 2
2 2 7 3 5 n E h 7 5 9 4 5 0 7 9 0 9
0 7 5 8 0
0 7 e 1 2 1 2 p C 0 0 0 1
0 1
0 0 1 1 2 2 1 2 1 p
0 0 0 0 M 0 0 0 A
6 5 5 6 6 5 6 6 6 6 6 5 6 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 a 0 0 0 0 0 0 0 0 0 0 0 0 0 5 0 0 0 0 0 0 0 0 0 0 0 0 0 O
. N 1 i i 1 i 1 i 1 i 1 i i 1 L E 8 9 5 5 1 2 5 8 2 8 9 9 2 1 d
K 8 1
0 0
0 0
9 2
9 0
0 7
9 5
9 2
9 0 8 4 3 s e 9 0 0 9 9 8 9 9 9 0 9 9 9 9 9 9 9 n t a 0 1 1 0 0 1 0 0 0 0 0 0 0 i
o l u
t c l
a l c
a 0 1
1 1
2 1
2 1
4 1
2 1
1 1
0 1
1 1
1 1
0 2 2 u a 0
0 0
0 0 0 0 0
0 0
0 0
0 0 0 0 1
0 1
0 1
0 l
c 4 0 0 0 0 0 0 0 0 0 0 0 0 0 P 0 0 0 0 0 a N i 1 1 i 1 1 i 1 i 0 0 C 1 i i i C 4 8 0 6 0 8 8 0 3 0 5 2 6 k M 6 9
0 0
1 0
5 9
8 9
7 9
8 9
2 0
5 9
1 9
3 9
6 3 r 9 0 0 9 9 9 9 0 9 9 9 9
9 0
0 a 0 1 1 0 0 0 0 0 0 n
i 1 0 0 1 1 h t .
h A r e i c 6 4
6 4
6 4 #.
6 4
6 4
6 4
6 4
6 4
6 4
6 4
6 6 4
4 r 4 e B E n 2 2 2 2 2 2 2 2 2 2 2 2 2 l
b y t
a i l
T a c
i t
i r
C f
o n i
o y t a _
r c a i f
i t
m n e
m d u I
" I S H V
" I I
I I
I _
H V V H X I 1 I V V I H I I X X X X X X X X X e
r e
r e
r e
r e e e e e e e e e r r r r r r r r r o o o o o o o o o o o o o C C C C C C C C C C C C C 3
) ) ) ) ) ) ) ) ) )
8
)
e 7 7) .
7 7 7 7) .
7 7 7 7 7 7 7 9 1
c n A A A A A A A A A A A A A 8 e 4
(
4
(
4
(
4 4 4 4 4 4 4 4 4 4 9-( ( ( ( ( ( ( ( ( (
r 4 4 I e 4 4 4 4 4 4 4 4 4 4 4 f
e 8
4 8
4 8
4 8
4 8
4 8
4 8
4 8
4 8
4 8
4 8
4 8
4 8
4 H
R 1 1
1 1
1 1
1 1
1 1
1 1
1 t
r W W W W W o W W W W W W W W p
& & & & & & & & & & & & & e B B B B B B B B B B B B B R c
t e
2 3 4 0 1 2 3 l 1 5 6 7 8 9 1 1 1 1 o H
a 0 1
5 1 1 6 0 0 1 _
O 1 0 3 8 6 3 8 e
) N 0 2
7 5 4 6 3 C C C C C C 1 0 g Ve F.
1 1 5 4 N N N N N N 1 a .
f K 0 0 0 1 1 0 0 P _
' A _
F a 4
L 4 3 4 x A P 0 2 4
4 5
7 3 1 1 3 2 8 7 6 d
i E N 2 1 7 1
5 1 4 4 4 5
4 2
4 3
5 9
4 7
1 7
7 7
6 3
7 C n C 0 0 0 0 1 4 2 1 1 1 N e p
1 1 0 0 0 0 0 M 0 p -
A _
5 6 6 6 6 5 6 a
0 0
0 0
0 0
0 0
0 0
0 0 0 0
0 0
0 0 0 0
0 -
5 0 0 0 0 O 0 0 0 N i 1 i i 1 i C C C C C C
._ E N N N N N N 1 k 1 2 8 4 3 9 4 K 7 9
3 9
1 9
2 9
1 9 9 4 0 0
s W- 9 9 9 9 9 9 0 n 0 0 0 0 0 0 i
o 8
=
1 t l c
a a 2 1 0 0 0 9 0 9 3 2 0 0 l
C 1 1 1 1 1 0 1 0 1 u a 0
0 0
0 0
0 0
0 0
0 0
0 0 0 0 1
0 1
0 1
0 l
c 4 0 0 0 0 0 0 0 0 0 0 P 0 0 0 0 0 0 a N 1 i i 1 1 i i 1 i i 0 0 C
C C 1 8 0 0 2 3 0 6 2 i i 9
N k M 6 9
0 0
N 9
7 9
9 9
7 9
2 0
6 0
6 5 3
4 7 r 9 9 9 9 9 0 9 9 9 9 0 0 9 9 9 9 a 0 1 0 0 0 0 1 1 0 0 0 0 m
1 h c . 2 0
A n h c 4 4 6 6 4 4 4 5 5 5 5 5 4 e i r 4 4 4
/ 7 7 7 7 3 -
e B E n 2 2 2 2 2 2 2 6 4 4 4 4 4 2 .
l b y 2 _
t a i l
T a c
i t
i r p C B a g
p a
p a n f B B m g g o o n p p m i
t o
m m p a c m a r
y i t
p p g m c a r a p p 6 5 c 0 p c 5 B 0 2 5 a i f
i 6 6 1 B 1 e s
t 8 4 1 m r r r r O m n e
8/ 7
/
/
w m
p p t o t o t o t e ,
p p a a a a r m d w w ,
r r r r o u
l I , , e 7 9 e e e e t l
e l
u 3 9 d d d d c S X I u
e u F 3 8 1
o o o o l e
I X X F F e 1 2 M M M M f e
X X X e e p y 1 1 r r r r R e e e p p t e e e e e e l r r r y y - s t t t t e o o o t t
- 0 a s
a a a a a e C C C S S S C C W W W W S t
3
) ) ) ) )
8 e
- 7) .
7 7 8 8 8 9) .
- 9) .
)
1 9 1
1 c
n A A A A A A A A )
0
)
0
)
0
)
0 8 e 4
(
4
(
4
(
4
(
4
(
4
(
4
(
4
( 1 1 1 1 A 9-r 4
( I f
e 4 8
4 8
4 8
5 4
5 4
5 4
0 1
0 1
A A A A 2 H e 4 4 4 6 6 6 8 8 4 4 4 4
( ( ( ( 0 R 1 1
1 1
1 1
1 1
- h h h h 6 tr W W c c c c 3- o W W W W W W n e
n e
n n L p
& & & & & & & & r r e
r e
r N e B B B B B B B B F F F F P R c
e t
4 5 6 7 8 9 0 1 2 3 4 5 6 l 1 1 1 1 1 1 2 2 2 2 2 2 2 o H
a 1 5 9 5 0 8 2 9 7 0 4 0 1
O 0 7 7 6 9 4 0 9 9 9 9 8 3 2 6 6 5 5 4 3 e
) N 0 0 0 0 2
3 0
3 9
2 8 8 1 1 0 9 g Ve E 0 0 0 0 0 0 0 2 2 3 3 3 2 P
a
( K 0 0 0 0 0 0 ,
' A F a 4 L 4 0 x A P 0 1
8 6
7 3
7 6
1 1 8 1 5 0 3 i d
E N 0 9 9 9 C 0 9 1 2 8
5 8
3 1 C 3 0
8 8 n C
1 0 0 0 N 3 2 2 2 3 N 3 2 e 0 0 0 0 0 0 0 0 p M 0 0 0 p .
A _
6 6 6 6 7 7 7 7 7 7 7 7 7 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 a 0 0 0 0 0 0 0 0 0 0 0 0 0 5 0 0 0 0 0 0 0 0 0 0 0 O 0 0 N i i 1 i i 1 1 1 1 1 1 i L E 2 4 0 9 3 2 4 1 7 0 3 7 5 I
t K 9 9
6 9
0 9
3 9
0 0
1 0
7 9
5 9
4 7 0 9 8 s n 9 9 9 9 9 9 0 9 9 a 0 0 9 9 9 9 0 9 9 n " 0 0 0 0 1 1 0 0 0 0 i
o '
1 0 0 t l e _
a a 9 9 0 8 0 2 1 0 1 1 2 l
C 0 0 1 0 1 1 2 u
1 1 1 1 0 0 0 0 0 0 0 0 0 0 0 c a 0 0 0 0 0 0 0 0 0 0 0 l 4 0 0 0 0 P 0 0 0 0 0 0 0 _
a N 1 i i i C C C N i i i i i N i 1 C 0 8 4 2 7 4 9 0 1 5 0 k M 8 9
6 9
7 9
6 9
9 9
9 9
6 9
1 9
4 9
2 0 r 9 9 9 9 9 9 9 9 9 0
0 0
0 a 0 0 0 0 0 0 0 0 0 1 1 m
1 h c .
A n e
h i
c 5 3 3 5 5 3
5 3
6 0
6 0
6 0
6 0
6 0
6 0
6 0
6 0
6 0
4 r 3 3 3 3 3 3 3 3 3 e B E n 2 2 2 2 4 4 4 4 4 4 4 4 4 l
b y t
a i l
T a c
i t s i
r n n n n
e n
".n t
e e
n n
C p e
s p
e s
p e
s n
p p
e p
e p
e p h
S n
. p e
p e
f .
s s s l s n e n e a p s o o m c
m c
m c
s e
p e m m m s e
r o
e s m m y i s c c c c c t
a 1 6 2 t t B m r c 2 1 i
n m 1 6 5 i c 5 6 fm 1 2 0 ht a i f
i 3 6 9 i f
n c 3 1
6 4 n i 0 5
0 5
9 m t n 1 2 3 i O 1 2 5 I w 1 e r r r r
r r
r r
r r
r r
r m d o o o o o o o o o o o o o u I t t t t t t t t t t t t t c c c c c c c c c c c c c S l e l e l e l e l e l e l e l e l e l e l e l e l e
f f f f f f f f f f f f f e e e e e e e e e e e e e R R R R R R R R R R R R R l l l l l e e e e e l
e l
e l
e l
e l e d d d e e e e e e e e e e a a a t t t t t t t t t t s e e S S S S S S S S S S I L L 3 -
) ) ) ) ) ) ) ) ) ) )
8
) )
1 1 1 1 I 1 I 1 1 1 2 2 2 9 e 1 1 1 1 1 1 1 1 1 1 1 1 c I. I. 8 n
e A A A A A A A A A A A A A 9-r 4
(
4 4 4 4 4 4 4 4 4 4 4 4
( ( ( ( ( ( ( ( ( ( ( ( I e
f e 2 0
2 0
2 0
2 0
2 0
2 0
2 0
2 0
2 0
2 0
6 2
6 2
6 2
H R 6 3
6 3
6 3
6 3
6 3
6 3
6 3-6 6 6 9 9 9 t r
r e e e r 3
r 3 3 3 3 3 o i
N I
N b h
I N
I N
I N
L N
I N b b I b b p e
h h N h h P P P P P P P P P P P P P R .
c e
t 7 8 9 0 1 2 3 4 3 6 7 8 9 l 2 2 2 3 3 3 3 3 3 3 3 3 3 o H
a 2 1
5 4 9 4 4 4 2 3 1 2 O 5 5 5 6 6 6 7 6 8 e 7 1 2 6
) N 8 2
1 1
1 1 1 1 1 1 1 1 8 8 C C 8 g Ve E 1 1 1 1 3 3 N N 8 a f K 0 0 0 0 0 0 0 0 0 0 0 P
' A F a 4 L 4 x A T 1
3 5
5 2
7 1 5 2 2 3 9 1 i d
E N 8 1 C C C 1 6
1 6
1 2
7 7 4 9 8
0 5
7 1 n C 2 1 N N N 1 1 1 3 3 2 5 9 e 0 0 0 0 0 0 p M 0 0 0 0 p A
7 7 7 7 7 7 7 7 7 7 6 0
0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 a 0 0 0 0 0 0 0 0 0 0 0
0 5 0 0 0 0 0 0 0 0 0 O 0 0
. N 1 i i 1 1 1 i i 1 i C C 1 t E 6 0 1 5 2 1 2 9 6 3 N N i
K 4 5 7 6 7 8 8 6 6
9 9 9 9 5 6 4 d 9 9 9 9 9 9 0 s 9 9 9 9 9 9 9 9 9 9 0 n
- 0 0 0 0 0 0 0 0 0 0 i
o t
==
1 t l c -
a a 2 2 0 2 2 1 1 2 0 1 l C 1 1 1 1 1 u a 0
0 0
0 0 0 0 0 1 1 0
1 0
1 0 0 1
l c 4 0 0 0 0 0 0 0 0 0 0 P 0 0 0 0 0 0 0 0 a N C C C C C i
1 1
5 N N N 2 1
6 i
4 1 1 i i i i -
1 9 4 7 1 k M 7 9
2 9
8 9
9 9
9 9 9
6 7 5 4 r 9 9 9 9 9
9 9 9
9 9 0 0 a 0 0 0 0 0 0 0 9
0 0
1 1 0
m 1 h c .
u A n h c 6 0
6 6 6 6 6 6 6 6 6 6 6 P e i 0 0 0 0 0 0 0 0 0 0 0 4 r 3 3 3 3 3 3 3 3 3 3 3 %
e B E n 4 4 4 4 4 4 4 4 4 4 4 3
4 0 l
b y 2 t
a i l
T a c s s i r e e p t e t t p a i b s s a a a r r n r
o s t e e l p l p r t 1
C t
p s e a a l t
x 2 e b t l p l p e l
e x u f s a a e e u R o n m o l
p l e
l e
t S-t S- n h m o c n r e e s ht B E y i t t t n n e i t
r a 5 - Z S S o r o r
t a i w m 2 c 0 2 - - - a o l w - B p a i f
i 4 3 B B p -
3 p 3 5 0 l a m e m t n /
4 r R V p 0 p _
e r 0
0 3
3 1
4 % % o 4 4 p 5 y m d o 0 0 0 1
0 5 0
2 B
1 1 O
5 T
u I t c
e t
t n
t n
t n 1 6
2 t
2 t
2 S D e e e 1
n n - - l e
e e e 9 2 1 3 7 u R R i m
r i
m r
i m
r 0
0 1
0 3
0 i m
r i
m 7 1
7 1
F r
d a p o
p e
p e
p p
p
. e p e e e X
= p _
x s s O _
e x x x x x a a h x l E E E F E E E E C C M _
3
) ) ) ) ) ) ) ) ) )
8 2 ) ) ) 9 3 3 3 3 3 3 3 1 4 5 5 6 e 1 1 1 1 1 1 1 1 1 1 1 1 1 c 1 8
n e A A A A A A A A A A A A A 9-r 4 4 4 4 4 4 4 4 4 4 4 4 4
( ( ( ( ( ( ( ( ( (
_ e ( ( ( I f
e 6 2
5 1
5 1
5 1
5 1
5 1
5 1
5 1
7 6
7 6
7 6
7 6
3 0
H _
R 9 6 6 6 6 6 6 6 1 1 2 2 8 t r
3 2 2 2
- r 2
r 2
2 2 7 7 r
4 r
4 5 o b
h b h
1 N
I N
I N
1 N b b b I i b b p e
_ P P P P P P h
P h h N N h h R P P P P P P c
e 0 1 2 3 4 5 6 7 8 9 0 1 2 l t
4 4 4 4 4 4 4 4 4 4 5 5 5 o H
=
s u
m e 3 -
l 1 5 4 6 5 7 0 7 4 5 7 a O 4 0 6 1 0 9 C 5 5 4 t l e N 9 2
7 1
1 1
4 8
5 4
1 5 N 9
2 5
1 9
1 na e n g
a Me E 0 0 0 0 0 0 0 0 0 mo P
( K i r
i t
' el a A pu 4 f
l a xc x 4 8 5 9 5 6 9 el A P 6 6 3 6 7 0 9
0 3
2 0
2 6
3 ec a i d
E N 9 6 1 6 4 2 3 9 5 0 g n C 2 1 1 0 4 5 6 2 1 1 r e e p
0 0 0 0 8 0 0 0 0 0 ah t _
M l yg p w -
. l n A aui n
_ l i 6 6 6 6 7 6 7 6 6
- 0 0 0 0 0 0 0 0 0 s e 0 0 0 0
0 0
0 0
0 0 0 0 0 u m 0 0 0 0 0 n
ut e r
_ 5 0 0 8 8 0 0 0 0 0 O C e N i i i i i i N 1 1 1 f od k . EK 6 3
9 9
6 6
5 0
6 5
7 4
7 6
3 3
8 0 yin 0 9 9 0 9 0 9 1 0 t S > 0 9 9 0 9 0 9 0 0 i
l d n 1 0 0 1 0 1 0 1 1 i e b n o - i s
i st a i
t .
a
. 2 1 1 1 0 1 0 1 1 1 oe pr l
u c 1 0
1 0
1 0
1 0
1 0
1 0
1 0
1 0
1 0
1 c e 0 0 0 0 0 0 0 0 0 0 0
ee h r
_ 4 0 0 0 0 0 0 e l 8 8 0 a P N i i i i i i i i i 8
gw t
C C 8 3 6 0 6 4 i ny k M 5 0
8 0
M 0
9 9
0 3 0
8 0 9 3
6 9 3
i t
sh e
0 0 9 0 _
r 0 0 0 0 9 0 0 0 9 0 et _
g a 0 _
gm 1 1 1 9 1 1 1 0 1 1 m u s ts 1 h u u u )
i a
c .
u u u u 0 v
_ A n e
h c P P P P 4 7
P P 4 P 4 3 r 7 7
> es i
4 r % % % % % % %
B n
0 0 0 6 5 6 6 5 6 5 .
l e y E 2 2 2 6 6 6 6 n(
os on b t i r c _
a s e r i
T l
a si i l f t f o .
u
_ m i
c h h h go d ,
t c c c n i
r t t t
i i l e
3 3 2 i p
i p p s ad C 4 1 3 h
c 2
O h
c "
9 h
c uc at l i u _
f p
=_ p 2 t i
p "6 u t i
p 7 t
i p cs cx so i
x x 5 5 o a E F E 0 " 0 P
d " 0 " ytae gt y 2 6 9 r s e t
a 2 2 2 2 5 2 e 5 2 7 r
ha beby t
c O O a O a i f
i 3
e 3
e 3
e P u 0 U u r o
0 U P u 0 U
t e ob l m
t n p p p P h l y y y t m d e
T T T 2
5 2
5 6
5 6
5 6
5 9
7 9
7 e r ta I gai u - - - e s s e e s
e s
e s
e e a r et S l e
u l
e u
l e a a a a a s
a s
a .eps vpu h C C C C C C C d j F F n o
n o
n o
n o
n n m t ea asl d X X X o o e aet ut hl uo u t t t t t t t O O O x e e x x e
x e
x e
x e
x a
l M M M S S S S S S S l cf s e ao r y c
c yl e t gaht t
)
7
)
7
)
7 7) )
7 7) )
7 or ne nh 3
8 e
)
6
)
6
)
6 1
A 1
A 1
A 1 1 1 1 r n emu e g 9 c
1 1 1 A A A A o 1 n A A A 4
(
4
(
4
(
4
(
4
( 4
(
4
( f er i o 8 9-r e 4 4 4 5 5 5 5 5 5 5 sh ehpl t e
( ( ( 8 8 8 8 8 8 8 d t I f
e 3 0
3 0
3 0
3 3
3 3
3 3
3 3
3 3 3
3 3
ns aie xA H R 0 0 0 3- t sFe . . tr 5 5 P- P- P- P- P P- P- r s o e 5- r A A A A A A A C Lseoi p C C C C C N Ah r rsa I
i N N i
N C C e P P P W W W W W W W ETeb R s c e*t t t e
3 4 5 6 7 8 9 0 1 2 t l 5 5 5 5 5 5 5 6 6 6 o o N H
E i
l i
1 i
l Table 4A.2 COMPARISON OF MCNP4a AND KFNO5a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS Calculated k,,r i 10 Enrichment MCNP4a KENO 5a 3.0 0.8465 0.0011 0.8478 0.0004 3.5 0.8820 0.0011 0.8841 0.0004 3.75 0.9019 0.0011 0.8987 0.0004 4.0 0.9132 0.0010 0.9140 0.0004 4.2 0.9276 0.0011 0.9237 0.0004 4.5 0.9400 0.0011 0.9388 0.0004 i
' Based on the GE 8x8R fuel assembly.
i Holtec Repon HI-981983 Appendix 4A, Page 14
p .
i l
l
)
l \
l l
Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS j Ak MCNP4a Worth of Calculated EALF' Ref. Experiment Absorber k,,, (eV) 4A.13 PNL-2615 Boral Sheet 0.0139 0.9994 i 0.0012 0.1165 I 4A.7 B&W-1484 Core XX 0.0165 1.0008 i 0.0011 0.1724 4A.13 PNL-2615 1.62% Boron-steel 0.0165 0.9996 i 0.0012 0.1161 4A 7 B&W-1484 Core XIX 0.0202 0.9961 0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.0243 0.9994 i 0.0010 0.1544 4A.7 B&W-1484 Core XVII 0.0519 0.9962 i 0.0012 0.2083 4A.11 PNL-3602 Boral Sheet 0.0708 0.9941 0.0011 0.3135 4A.7 B&W-1484 Core XV 0.0786 0.9910 i 0.0011 0.2092 '
4A.7 B&W-1484 Core XVI 0.0845 0.9935 i 0.0010 0.1757 4A.7 B&W-1484 Core XIV 0.1575 0.9953 i 0.0011 0.2022 4A.7 B&W-1484 Core XIII 0.1738 1.0020 i 0.0011 0.1988 4A.14 PNL-7167 Expt 214R flux trap 0.1931 0.9991 0.0011 0.3722 3
1 i
j TALF is the energy of the average lethargy causing fission.
Holtec Report HI-981983 Appendix 4A, Page 15 1
i I
i l
4 l
I Table 4A.4 j
COMPARISON OF MCNP4a AND KENO 5a CALCULATED REACTIVITIES' FOR VARIOUS ' B LOADINGS ;
l Calculated k,,, la 3'B, g/cm 2 MCNP4a KENO 5a 0.005 1.0381 t 0.0012 1.0340 0.0004 l 0.010 0.9960 0.0010 0.9941 i 0.0004 1 0.015 0.9727 0.0009 0.9713 0.0004 0.020 0.9541 0.0012 0.9560 i 0.0004 0.025 0.9433 i 0.0011 0.9428 t 0.0004 0.03 0.9325 0.0011 0.9338 0.0004 0.035 0.9234 i 0.0011 0.9251 i 0.0004 0.04 0.9173 0.0011 0.9179 0.0004 Based on a 4.5% enriched GE 8x8R fuel assembly.
Holtec Report HI-981983 Appendix 4A, Page 16
g-I i
l l
l Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORS' Separation, Ref. Case E, wt% cm MCNP4a (, KENO 5a %,
4A.I1 Steel Reflector 2.35 1.321 0.9980 i 0.0009 0.9992 i 0.0006 2.35 2.616 0.9968 i 0.0009 0.9964 i 0.0006 2.35 3.912 0.9974 i 0.0010 0.9980 i 0.0006 2.35 =
0.9962 i 0.0008 0.9939 i 0.0006 4A.I1 Steel Reflector 4.306 1.321 0.9997t0.0010 1.0012 i 0.0007 4.306 2.616 0.9994 0.0012 0.9974t0.0007 4.306 3.405 0.9969 i 0.0011 0.9951 t0.0007 4.306 a 0.9910 i 0.0020 )ei i0.0007 4A.12 Lead Reflector 4.306 0.55 1.0025i0.00ll 0.9997 i 0.0007 4.306 1.956 1.0000 i 0.0012 0.9985 i 0.0007 4.306 5.405 0.9971 i 0.0012 0.9946 i 0.0007 Arranged in order of increasing reflector-fuel spacing.
Holtec Report HI-981983 Appendix 4A, Page 17
r l
i Table 4A.6 '
CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated k,,r Boron Concentration, Reference Experiment ppm MCNP4a KENO 5a 4A.15 PNL-4267 0' O.9974 0.0012 -
4A 8 B&W-1645 886 0.9970 i 0.0010 0.9924 i 0.0006 4A.9 B&W-1810 1337 1.0023 i 0.0010 -
4A.9 B&W-1810 1899 1.0060 t 0.0009 -
4A.15 PNL-4267 2550 1.0057 0.0010 -
l l
Holtec Report HI-981983 Appendix 4A, Page 18 l
Table 4A.7 i
CALCULATIONS FOR CRITICAL EXPERIMENTS WITII MOX FUEL MCNP4a KENO $a Reference Case' L EALF" h EALF" PNL-580', MOX Fuel - Exp. No. 21 1.0041 to.0011 0.9171 1.0046 i 0.0006 0.8868
[4A.16}
MOX Fuel - Exp. No. 43 1.0058 i 0.0012 0.2968 1.0036 1 0.0006 0.2944 MOX Fuel - Exp. No.13 1.0083 i 0.0011 0.1665 0.9989 i 0.0006 0.1706 MOX Fuel - Exp. No. 32 1.0079 i 0.0011 0.1139 0.9966 1 0.0006 0.1165 WCAP- Saxton @ 0.52" pitch 0.9996i0.0011 0.8665 1.0005 0.0006 0.8417 3385 54
[4A.17] Saxton @ 0.56" pitch 1.0036i0.00ll 0.5289 1.0047i0.0006 0.5197 l
Saxton @ 0.56" pitch borated 1.0008 i 0.0010 0.6389 NC NC Saxton @ 0.79" pitch 1.0063 i 0.0011 0.1520 1.0133 i 0.0006 0.1555 Note: NC stands for not calculated Arranged in order of increasing lattice spacing.
EALF is the energy of the average lethargy causing fission.
Holtec Report HI-981983 Appendix 4A, Page 19
2 -
g
- g _ -
, s i
a X B E D
N 3 - rI 1
9 f
oL 0 _
_ 8 n o A f
o _
i s S R 7 s E T _
f n
i F UC _
- , L E i
c ,
5 g
n AP VS i
i f
f o _
, s u -
o ,
5 a fE -
C C fH _
n O _ __ y eT o
_-'__ O g r
i t 4 o )
e kF -
a - _ _ 'h l O
l O O a e
r - _ _ t e S c D r
o L ES C _-
O e g
o TE
- }
_s
)
j s
. - 3 g L
( AU h
- .-' =
o a r L L UA t 6 ,
i w e O -
- A v CV L
n i
o -
)
f o AS s
s - r { -
j" O
y CU PI O e
r - T s - - _- -
2 g
R g
e _
O
- o __3
_t r 1 We' _u 9 E
r e
n NR CA r _
O
_- 7
_ MV a _ _- O_- 3 -
e -
0 -
i n _
L 1
_- L ,
, _ ~
~ -
0 - A
_- J _
- - 4 O _ g
_ . [ " ' E
__ 3 " R
_ _ - _ _ - ~ - _ __ _ _- , ~ _~ _ _ _~ 9 U
0 5 0 5 0 G I
1 0 0 9 9 F 0 0 0 9 9 1 1 1 0 0 et. g t eix. g j _ .
-O
2 TJ_
s '
i a ' X B ' E RD ON FI 1 '
9 2
S L
~
0 ~
8 n E RA o
f o ' i s U T
~
7 i s L C f
n
~
F A i
- ~
~
6 g V E P i
c ~ n S i
f f
f ~ s f o u a eE C
c 1 5 C
kT H
n - T_
y g
o i
t 4 r
a )
e DF l
a I h la EO a
r t
e S c T r
o L AS g LE C _
p 1~ 3 e
g L
(
o UU h
t i
(
' Ai a r
e CL LA w
A v AV n _
C c f o S aU i
2 - ) 1 y 5 O N -( 2 9_
2 3__ T*{- g r r 0I N R g e R _
T4 " x E n E A r _ K V o _
i s
n _
I 2 L _
A
_ TL - 4
~ E
) . ET_ R
- _ :f 2 E: ~
3 U
- - : : - : : - - _J _ ~
- G 0 5 0 5 0 5 I 1 0 0 9 9 8 F 0 0 0 9* 9 9 1 1 1 0 0 0
. e %e ;g3o oO
0 -
i i
6 .
i i
o __OO_
i i
- i i
- i 5
- i 5
i
- i i _
- i i
- i i _
3 i 0 0 i5 S
i 0 - i _
- ,. - L _
i T _
f o - 4 -
- J -
i i S N
_ i E E U
L M i
t n _
e i 5 i
c
_ i4 AH i
f
==" h, - O ;; - ,
r i
i VC I f
e o __ __ o ==' o _a 1 i
i R _
f o
i f N C i i eE n 0 i
o _ i45 k 5 t
a _
i i
3 3 2
l e _
i D 2 r
r i
i U E - _
o _ i TU C i o
A t
h i
ii35. w/
i L S UU -
C LI O 2
w i i
t n o - i o - -
i n e
AR i
s
- i i CA s - m V e i 0. h P r _ 3 cr g i i
n NT R
e s
_ i i
E CA r .i a _ i
, M a _ i e
B i n T s 5 3
j:_ ~ hJt T i
_. ). '::-_l , '
fI_ _
L ,
i i 2
_ 4_ 5l uhN3~ ~
E i A
_ - i i
4 i
_ i i
E
_ R i 0
_ _ _ _ _ - _ ~ _~ ___~:_ : __I _~ ~[ __~_~ 2 U 5 0 G 0 5 0 I 1 0 0 9 9 F 0 0 0 9 9 1 1 1 0 0
,>_oO ; ix
. - me o ) j
0 6
- o. __
U
- , 5
- , 5 s ,
i a ,
' B ,
~ ,
~ ,
- ,0 8
3 - ,5 0 - , S T
f o
, SN EE UM t ,
n - ,5
- L H A
i c - ,4 VIC i
f ,- _
f o
oEg h_c_ == __ __
, R o ,
fN C - f eE n - ,0 o
i t
- ,45 k 5 a ,
3 3 t
2 2 s
D-r _ , -
r ,
U E o _ ,
TU C ,
o A h
t
,,3 5. / LS U U i .
w w _
- i i
i e
i s
i i
m CA s
e
- i
- 0. hc V r
g - ,3 O R
e -
i i
r n N AT i
E E K
i r i a i e -
i n i 5 4
1;,
i r !=
. V0 y_e . r vd= LE-:_ gO_ : o L ,2 0 LC_ W _-)
O ,
A _
4 _
E
,0 R 2 U 0 5 0 5 G _
0 5 I _
1 0 0 9 9 8 F _
0 0 0 9 9 9 _
1 1 1 0 0 0
- 5e elx o =3_3 _
oU
0.94 _
0; 4.5s. E r
L
{ ) 4.2m E 0.92 <
N.05 E r
~
y -
o 0.90 - 'O 3.755 E 3
.2 -
O o -
o 3.!sE
% 0.88 7~
i 2
0 -
n -
o z -
w _
M 0.86 ,
d 3.Os E
~
0.84 iii, ,,,, .. ,,,, ,,,, ,,,, iii, iiii . . ii , iii.
0.84 0.86 0.88 0.90 0.92 0.94 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KEN 05A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS
r.
1.04 _ g 1.03 _
1 nns g/mn .n q{
1.02 _
~
1.01 i Q. _ !
Z -
~
U 1.00 _
y __
j) 3. 010 g /c meg h 0.99 R _
~
] 0.98 _
~
c _
3 0.97
~
O -
O -
~
i 0.96 x : _/ !
g : j_o e.e2e ./... .
I
- - 0.95 _ !
O _ _
h j_ O.025 i/cmeq e 0 ~. 94
~
"' ~ " !
0.93
_/
/-)- 0.025 g /c me a 0.92 1
-[
j 0.e4 . . ... l
~
0.91 i i i i i iii i i i i iiii i i i i i i i e i iii !
0.900 0.920 0.940 0.960 0.980 1.000 1.020 1.0 j Reactivity Calculated with KEN 05a FIGURE 4A.6 COMPARISON OF MCNP AND KEN 05a CALCULATIONS FOR VARIOUS BORON-101 AREAL DENSITIES
I' t
L l 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction l This section provides a summary of the methods, models, analyses and numerical results to demonstrate the compliance of the reracked Oyster Creek spent fuel pool with the provisions of Section III of the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications", (April 14, 1978).
Similar methods of thermal-hydraulic analysis have been used in other rer ick licensing projects (see Table 5.1.1).
The thermal-hydraulic qualification analyses for the rack array may be broken down into the following categories:
- i. Evaluation of the long-term decay heat load, which is the cumulative spent fuel decay heat generation from all fuel assemblies stored in the pool from previous discharges to the final discharge.
ii. Evaluation of the decay heat load and pool bulk temperature as a function of time -
during the postulated firal discharge scenarios.
iii. Evaluation of the " time-to-boil" if all forced heat rejection paths from the pool are L lost.
iv. Determination of maximum local water temperature in the fuel rack cells at the I instant when the bulk temperature reaches its maximum value.
! v. Evaluation of peak fuel cladding temperature in the hottest fuel cell.
l l The following sections present the plant system description, analysis assumptions, a synopsis of the analysis methods employed, and final results.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-1 Report HI-981983 l
o 5.2 Soent Fuel Pool and Cooline System Descriptions The spent fuel storage pool is 27 feet by 39 feet in plan with a total water depth of approximately 37 feet 9 inches, and an actual physical depth of 38'-9". The depth of the water to the top of the stored fuel is approximately 25 feet, providing approximately 200,000 gallons of water above the fuel.
To avoid unintentional draining of the pool, there are no penetrations that would permit the pool to be drained below one foot above the active fuel. All lines extending below this level are equipped with suitable valving to prevent backflow. The' passage between the fuel storage pool and the refueling cavity above the reactor vessel is provided with two double sealed gates with a monitored drain between the gates. This arrangement permits detection of leaks from tne passage and repair of the gates in the event of such leakage. A liquid level switch monitoring pool water level is prosided to detect loss of water and permit refilling of the pool from the condensate transfer system. This switch alarms in the Control Room. In addition, a second level switch in the pool surge tank is provided to permit almost instantaneous water loss detection. This detector alarms locally and in the Control Room. A low-low level switch is provided to automatically trip the SFPCS pumps upon reaching set point. Radiation monitors on the operating floor near the
' spent fuel pool alarm on high radiation and initiate isolation of the Reactor Building ventilation and operation of the Standby Gas Treatment System.
The fuel pool cooling system cools, filters, and demineralizes the fuel pool water. Failure of the fuel pool cooling system cannot cause the fuel to be uncovered. Normal demineralized water makeup to the pool is provided from the 525,000 gallon (nominal capacity) Condensate Storage Tank at a rate of 250 gpm by a single condensate transfer pump. The makeup capability from this system is increased to about 420 gpm if both condensate transfer pumps are used. Makeup water is added directly to the pool's surge tank by manual valve operation on El.119'. Additional makeup, at a rate.of 150 gpm, can be provided from the 30,000 gallon (nominal capacity)
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-2 Report HI-981983
l l
l Demineralized Water Storage Tank by the demineralized water transfer pumps through the use of hoses. Other sources of water are also available through the use of fire hoses or portable pumps.
The makeup system for the spent fuel pool is not a Seismic Category I system. The 2,000 gpm I
diesel driven fire pumps for the Fire Protection System can be used to provide makeup water from l the Fire Pond to the Condensate Storage Tank through a permanent connection.
! The Spent Fuel Pool Cooling System (SFPCS) is comprised of two spent fuel pool cooling pumps, l two spent fuel pool shell-and-tube heat exchangers, two augmented fuel pool cooling puraps, one plate and frame augmented fuel pool cooling heat exchanger, a filter, a demineralizer, two surge tanks, associated piping and valves, and interconnections to the condensate demineralizers and the condensate systems. The fuel pool cooling pumps and heat exchangers operate continuously to circulate the pool water and maintain its temperature at or below 125*F.
The fuel pool heat exchangers are located near the pumps. They are stacked one above the other L in saddle mounts. The cooling capacity of both heat exchangers in parallel, with the flow from I
both pumps,'is approximately 5.5 millicn Bru/hr at design conditions. Normally both heat exchangers are valved in for service, with the single interconnecting valve open at the inlet lines j so either pump can supply both heat exchangers. RBCCW System flow is adjusted to each heat
! exchanger by butterfly valves. Fuel pool water outlet temperatures from each heat exchanger are recorded, and high temperature is annunciated.
The two spent fuel pool heat exchangers are shell-and-tube construction, each designed to remove 2.75 million Btu /hr of decay heat from the pool while it is maintained at 125'F with RBCCW water supply at 90*F and 500 gpm flow rate. The decay heat removal capacity at the Oyster Creek j spent fuel pool is significantly enhanced by a large capacity plate and frame exchanger designated i l as the augmented fuel pool heat exchanger. The 800 gpm fuel pool water flow through this exchanger is limited by the size of the fuel pool weirs to prevent flooding. Each augmented fuel pool pump is rated at 1200 gpra and one pump is operated to meet the operational needs. The SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-3 Report HI-981983 <
decay heat removal capacity of the augmented fuel pool heat exchanger is 12.2 million Btu /hr at
^
the 800 gpm fuel pool water flow rate when supplied with RBCCW water at 90*F and 2000 gym flow and the pool maintained at L5 F.
5.3 Decav Heat Load Calen1ations The decay heat load calculation is conservatively performed in accordance with the provisions of USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July,1981.
To determine the end-of-life decay heat in the Oyster Creek spent fuel pool, the historical and projected discharges are considered as shown in Table 5.3.1. A total of 2,608 assemblies will be accumulated to the end of Cycle 18 from previous discharges. A full core discharge of 560 assemblies at the end of Cycle 18 (last operating cycle) will increase the fuel inventory to 3,168 assemblies, which conservatively exceeds the 3,035 storage locations in the pool.
A bounding long-term decay heat calculated from the discharge plan will be considered for the analysis. Since the 4ay heat lord from the old assemblies varies very slowly as a function of time, the long-term decay heat is assumed to be constant. The decay heat evaluation considers 2,053 days of in-core irradiation at full reactor power so as to conservatively bound decay heat in the Oyster Creek spent fuel pool. The decay heat calculation method is conservative with respect to other ORIGEN based calculation methods, which have been reviewed and accepted in other dockets. Therefore, the SRP 9.1.3 stipulated uncertainty for a relatively smaller number of old fuel assemblies accumulated in a fuel pool is not used.
i 1
l l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-4 Report HI-981983
r 5.4 Discharge Scenarios A total of three discharge scenarios are considered at the end of the final cycle of plant operations.
'Two cases for a partial core discharge are analyzed. One case for a full core offload (Case iii) is i analyzed. The two partial core discharge scenarios are based on normal cooling with the large l capacity-augmented heat exchanger (Normal Case ii) and with the smaller capacity shell-and-rube heat exchanger (Abnormal Case i) (in the event that both pumps of the augmented heat exchanger l
- are not available).
The following discharge scenarios are considered for bulk pool thermal-hydraulic analysis:
Case (i): An abnormal partial core is transferred to the spent fuel pool six days after reactor shutdown. The average rate of fuel transfer is 5.8 assemblies per hour. The heat from this fresh batch and background heat from old ,
discharges is removed by the smaller capacity shell-and-tube heat exchanger l with one pump. Consistent with the SRP 9.1.3 stipulation for abnormal scenarios, the bulk pool temperature shall be below boiling for this scenario.
Case (ii): A normal refueling cycle batch is transferred to the spent fuel pool six days after reactor shutdown. The average rate of fuel transfer is 5.8 assemblies per hour. The heat from this fresh batch and background heat from old discharges is removed by the augmented fuel pool heat exchanger with one pump in operation. The bulk pool temperature shall be maintained below 125'F for this scenario.
Case (iii): A full core is transferred to the spent fuel pool thirty-one days after reactor shutdown. The average rate of fuel transfer is 5.8 assemblies per hour. The heat from this full core and background heat from old discharges is iemoved by the augmented fuel pool heat exchanger with one pump in operation. The bulk pool temperature shall be maintained below 125 F for this scenario.
These scenarios are schematically depicted in Figures 5.4.1 and 5.4.2. Key thermal-hydraulic data for all cases is summarized in Tables 5.4.1 and 5.4.2.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-5 Report HI-981983
. 5.5 Bulk Pool Temperatures In this section, we present the methodology for calculating the bulk pool temperature as a function of the time coordinate. The method used to calculate the rate of pool water temperature rise and the time-to-boil, when all forced cooling paths are unavailable, is also presented.
In order to perform a conservative analysis, the heat exchangers are assumed to be fouled to their design basis level. Thus, the temperature effectiveness, p, for the heat exchangers utilized in the i analysis is the lowest postulated value based on the exchanger specification sheets.
1 The following assumptions also apply for the bulk pool temperature evaluation.
All old spent fuel assemblies discharged into the pool are assumed to have 2,053 effective Full Power Days (EFPD) of operation for all cycles. This assumption will provide a conservative decay heat calculation.
In calculating the spent fuel pool evaporation heat losses, the building housing the spent fuel pool is assumed to have the maximum ambient air temperature of 104 F and 100%
relative humidity. This assumption will result in conservative credit for the evaporation heat losses.
Heat conduction losses through the pool walls and slab are conservatively neglected, ,
The mathematical formulation can be explained with reference to the simplified heat exchanger
' alignment of Figure 5.5.1. Referring to the spent fuel pool cooling system, the governing differential equation can be written by utilizing conservation of erlergy:
C dsdT = Qt - Onx (5.5.1)
Qt = P + Q(t) - Grv (T,t,)
l L
SHADED TEXT CONTAINS PROPRIETARY INFORMATION f Holtec International 5-6 Report HI-981983 1'
L l L I
l :
\ l l
i- -
j where:
C: Thermal capacity of the pool (net water volume times water density and 1 times heat capacity), Btu /*F Qt: Net heat load in the SFP, Btu /hr.
Q(t): Heat generation rate from recently discharged fuel, which is a specified function of time, r, Btu /hr.
P: Heat generation rate from "old" fuel, Btu /hr. It is also termed as long-term
' decay heat load. The SFP pump heat is included in P for maximum bulk
)
j' l
pool temperature calculation.
Qsx: Heat removal rate by the heat exchanger, Btu /hr.
Qsv (T,t,): Heat loss to the surroundings, which is a function of pool temperature T and ambient temperature t,, Btu /hr.
Qnx is a nonlinear function of time and can be written in terms of effectiveness p as follows:
Qux = W, C, p (T, - t i) (5.5.2) to - t; P=T-t i where:
W,: Coolant flow rate, Ib./hr.
C,: Coolant specific heat, Btu /lb. F.
p: Temperature effectiveness of heat exchanger.
T:i Pool water temperature, *F t,: Coolant inlet temperature, "F t:
o Coolant outlet temperature, *F SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-7 Report HI-981983 i
t
\
l l-1 The temperature effectiveness p is calculated from data on the exchanger specification sheets. Q(t) is deterinined according to the provisions of USNRC Branch Technical Position ASB9-2, l
" Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July,1981. )
Q(t) is a function of decay time, number of assemblies, and in-core exposure time. During the l I
fuel transfer, the heat load in the pool will increase with respect to the rate of fuel transfer and
]
equals Q(t) after the fuel transfer, l Qev is a nonlinear function of pool temperature and ambient temperature. Qay contains the heat evaporation loss through the pool surface, natural convection from the pool surface and heat conduction through the pool walls and slab. Experiments show that the heat conduction takes only
]
l about 4% of the total heat loss (5.5.1], and can therefore be ~ conservatively neglected. The evaporation heat loss and natural convection heat loss can be expressed as:
l Qev = m C A, + ch sh 0 (5.5.3) l where:
m: Mass evaporation rate, Ib./hr. ft.2 l- P: Latent heat of pool water, Btu /lb.
l A,: Pool surface area, ft.2 h:e Convection heat transfer coefficient at pool surface, Btu /ft.2 hr.*F 0 = T-t : , The temperature difference between pool water and ambient air, F
! The mass evaporation rate m can be obtained as a nonlinear function of 6. We, therefore, have m=hp (0) (W,, - W,) (5.5.4)
W p,: Humidity ratio of saturated moist air at pool water temperature T.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-8 Report HI-981983
~
W,,: Humidity ratio of saturated moist air at ambient temperature t, lio(0): Diffusion coefficient at pool water surface. hois a nonlinear function of 6, Ib./hr. ft 2 op The nonlinear single order differential equation (5.5.1) is solved using Holtec's QA-validated numerical integration code "ONEPOOL".
The next step in the analysis is to determine the time-to-boil if all forced cooling paths become unavailable. Clearly, the most critical instant of loss-of-cooling is when pool water temperature has reached its maximum value. It is assumed that makeup water is added at the rate of G lb./hr.
The makeup water is at temperature tg. The governing enthalpy balance equation for this condition can be written as
[C + C'] S = ?, + Q (t + Tg) + G (tg - D - Ory (5.5.5) ds
'* TINS where C'
- G (T - T) - [0 Ogy. In the foregoing, water is assumed to have specific heat of unity and latent heat L, and the time coordinate T is measured from the instant maximum pool water temperature is reached. r, is the time coordinate when the makeup water application is begun. r,, is the time coordinate measured from the instant of reactor shutdown to when maximum pool water temperature is react.ed. T is the dependent variable (pool water temperature). For conservatism, Qey is assumed to remain constant after pool water temperature reaches and rises above 170 F. The time-to-boil calculations are conservatively performed assuming no makeup water is available.
A QA-validated numerical quadrature code is used to integrate the foregoing equation. The pool water heat up rate, time-to-boil, and subsequent water pool depth time profile are generated and j reported in this chapter.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-9 Report HI-981983
I l
1 5.6 Local Pool Water Temnerature l In this section, a summary of the methodology for evaluating the local pool water temperature is presented.
)
i 5.6.1 Basis The local water temperature analysis uses bounding bulk pool conditions coincident with peak bulk !
temperature response. In order to determine an upper bound on the maximum local water !
temperature, a series of conservative assumptions are made. The most important assumptions are listed below ,
1 Forced cooling of the bottom plenum by the existing sparger lines is conservatively neglected.
With a full core discharged into the fuel pool, the peak bulk pool temperature instant is evaluated as a steady state condition for local temperature analysis.
The nominal east and west wall downcomer gaps are considered in the local temperature analysis. The downcomer gaps along the north and south walls are conservatively neglected.
No downcomer flow is assumed to exist between the rack modules.
- The hydraulically limiting fuel assembly type is considered in the analysis.
5.6.2 Local Temperature Evaluation Method Local temperature evaluation of the Oyster Creek pool is performed under a conservatively postulated sparger line truncation condition. Adequate cooling of hot fuel in the fuel pool is demonstrated by performing a rigorous evaluation of the velocity and temperature fields in the pool created by the interaction of buoyancy driven flows and water injection / egress. A Computational Fluid Dynamics (CFD) analysis for this demonstration is implemented. The SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-10 Report HI-981983 1
objective of this study is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled conditions in the pool is met for the maximum fuel pool decay heat scenario. The l
local thermal-hydraulic analysis also considers the effect of a hot fuel assembly dropped above ;
the racks and laying horizontally on top of the hot fuel rack cells coincident with a 50% fuel cell exit blockage under this postulated mechanical accident scenario. An outline of the CFD approach is described in the following.
There are several significant geometric and thermal-hydraulic features of the Oyster Creek spent fuel pool which need to be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there are two regions to be considered. One region is the bulk pool region where the l
classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat generating fuel assemblies located in the spent fuel racks located near the bottom of the spent fuel pool. In this region, water flow is directed vertically upwards due to buoyancy forces through relatively small flow channels formed by the fuel assembly rod arrays in each rack cell.
This situation shall be modeled as a porous solid region in which fluid flow is governed by the classical Darcy's Law:
aP =-
V' - CplVjVi /2 a X, K(i) where ap/6Xi is the pressure gradient, K(i), Vi and C are the corresponding permeability, velocity and inertial resistance parameters and is the fluid viscosity. The permeability and inertial resistance parameters for the rack cells loaded with Oyster Creek fuel is determined based on friction factor correlations for laminar flow conditions typically encountered due to low buoyancy induced velocities and small size of the flow channels.
. The Oyster Creek pool geometry requires an adequate portrayal of large scale and small scale features, spatially distributed heat sources in the spent fuel racks and water inlet / outlet configuration. Relatively cooler bulk pool water normally flows down through the narrow fuel i
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-11 Report HI-981983 1
rack outline to pool wall liner clearance known as the downcomer. Near the bottom of the racks, the flow turns from a vertical to horizontal direction into the bottom plenum supplying cooling water to the rack cells. Heated water issuing out of the top of the racks mixes with the bulk pool water. An adequate modeling of these features on the CFD program involves meshing the large scale bulk pool region and small scale downcomer and bottom plenum regions with sufficient number of computational cells to capture the bulk and local features of the flow field.
The distributed t. at sources in the spent fuel pool racks are modeled by identifying distinct heat generation zones considering full-core discharge, peaking effects, and presence of background decay heat from old discharges. Three heat generating zones were modeled. The first consists of l
background fuel from previous discharges, the remaining two zones consist of fuel from a full- I core-discharge scenario. The two full core discharge zones are differentiated by one zone with higher than average decay heat generation and the other with less than average decay heat I generation. This is a conservative model, since all of the fuel with higher than average decay heat is postulated to be placed in a contiguous area. The analysis has been performed for a limiting full-core offload scenario selected from the offload scenarios listed earlier.
The CFD analysis is performed on the industry standard FLUENT [5.6.4] fluid flow and heat transfer modeling program. The FLUENT code enables buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying "Reynolds's Stresses" to the mean bulk flow quantities with the following turbulence modeling options:
- i. K-E Model ii. RNG K-E Model lii. Reynolds Stress Model The K-E Model is considered most appropriate for the Oyster Creek CFD analysis.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-12 Report HI-981983 1
Rigorous modeling of fluid flow problems requires a solution to the classical Navier-Stokes equations of fluid motion [5.6.1]. The governing equations (in modified form for turbulent flows with buoyancy effects included) can be written as:
1 >
ap,u, 89, u, u, a a u, a u,
~
at cX, BX, aX, BXi ,
O E - p, p (T - T,) gg + ' I OXi OX) where u, are the three time-averaged velocity components. p (u'i u[) are time-averaged Reynolds stresses derived from the turbulence induced fluctuating velocity componentsi u', p, is the fluid density at temperature T , p is the coefficient of thermal expansion, y is the fluid viscosity, gi are the components of gravitational acceleration and x, are the Cartesian coordinate directions. The l
Reynolds stress tensor is expressed in terms of the mean flow quantities by defining a turbulent viscosity p, and a turbulent velocity scale k* as shown below [5.6.2]:
p ( ui u[-) = 2/3p k 5'.l -
Bu' + au)
BX; SX, The procedure to obtain the turbulent viscosity and velocity length scales involves a solution of two additional transport equations for kinetic energy (k) and rate of energy dissipation (c). This methodology is known as the k-c model for turbulent flows as described by Launder and Spalding
[5.6.3].
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION ;
Holtec International 5-13 Report HI-981983 l l
5.7 Cladding Temperature In this section, the method to calculate the temperature of the fuel cladding is presented.
The maximum specific power of a fuel array q, can be given by:
I 1
4 9A =4 fxy (5.7.1) s where:
Fy = radial peaking factor q =
average fuel assembly specific power The peaking factors are given in Table 5.4.1. The maximum' temperature rise of pool water in the most disadvantageously placed fuel assembly, defined as one which is subject to the highest local pool water temperature, is computed for the maximum decay heat load full core discharge scenario. Having determined the maximum local water temperature in the pool, it is now possible to determine the maximum fuel cladding temperature. A fuel rod can produce F, times the average heat emission rate over a small length, where F, is the axial rod peaking factor. The axial heat I
distribution in a rod is generally a maximum in the central region, and tapers off at its two i extremities. Thus, peak cladding heat flux is given by the equation: '
qF y F, 4e "
A, where A, is the total cladding external beat transfer area in the active fuel length region.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-14 Report HI-981983
n
)
i I
Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially u'pwards from bottom to top under laminar flow conditions. Rohsenow and Hartnett [5.7.1]
l report Nusselt-number based heat transfer correlation for laminar flow in a heated channel. The film temperature driving force (ATr) at the peak cladding flux location is calculated as follows:
h, Dg/Kw = Nu q
AT=f r (5.7.4) where, hr is the waterside film heat transfer coeffici .nt, Dn is sub-channel hydraulic diameter, Kw is water thermal conductivity and Nu is Nusselt number from heat transfer correlation.
In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit resistance R, (equal to 0.0005 2ft -hr- F/Bru), which covers the entire surface. Thus, including the temperature drop across the crud resistance, the cladding to water I local temperature difference (AT,) is given by:
AT, = AT, + R, qc (5.7.5) 5.8 Results This section contains results culled from the analyses performed for each of the three postulated discharge scenarios.
s-SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-15 Report HI-981983 l
e.
5.8.1 ' Bulk Pool Temnerature The bulk pool temperature and the decay heat load profile in the spent fuel pool as a function of time (after reactor shutdown) are shown in Figures 5.8.1 through 5.8.5 for the three discharge scenarios postulated in Section 5.4. The maximum values of the pool bulk temperature and the coincident heat load to the spent fuel pool cooler are summarized in Table 5.8.1. As would be expected from physical considerations, the thermal inertia of the pool water causes the bulk pool temperature to reach its maximum value within a short time after the occurrence of the peak decay heat load, the lag is a direct result of the system thermal capacitance. 'Ibe fact that the coincident heat load is not the maximum decay heat load (due to aforementioned system inertia) warrants emphasis; since this distinction is often neglected in evaluation of the system performance data.
The coincident time to the maximum temperature is the summation of in-core hold time, fuel transfer time, and the lag time due to thermal inertia. It is shown from the results that the pool water temperatures are kept below 125*F during normal refueling offload (Case (ii)) and full-core offload (Case (iii)) discharge scenarios. For the Case (i) scenario, the peak bulk pool temperature is below boiling by a substantial margin.
The maximum bulk temperatures tabulated in Table 5.8.1 demonstrate the compliance of the Oyster Creek cooling system with the latest USNRC acceptance criteria. Therefore, no physical modification of the spent fuel pool cooling and purification system is necessary.
5.8.2 Time-to-Boil If all heat exchanger assisted forced pool cooling becomes unavailable, then the pool water will begin to rise in temperature and eventually will reach the normal bulk boiling temperature at 212*F. The time to reach the boiling point will be the shortest when the loss of forced cooling occurs at the point in time when the pool bulk temperature is at its maximum calculated value.
Although the probability of the loss-of-cooling event coinciding with the instant when the pool water has reached its peak value is extremely remote, the calculations are performed under this SHADED TEXT CONTAINS PROPRIETARY INFORMATION !
Holtec International 5-16 Report HI-981983 l l
extremely unlikely scenario. Table 5,8.2 contains the results with the additional proviso that no makeup' water was added to the pool. Figure 5.8.6 shows the variations of SFP water elevation L as a function of time after loss of cooling for all the analyzed scenarios. The time-to-boil results for the Oyster Creek pool are comparable to other BWR pools with densified fuel storage. !
l 5.8.3 Local Water and Cladding Temocrature Consistent with our approach to make the most pessimistic assessments of temperature, the local water temperature calculations are performed when the pool is at its peak bulk temperature. Thus, l the local water temperature evaluation is a calculation of the temperature increment over the theoretical spatially uniform value due to local hot spots (due to the presence of a highly heat i emissive fuel bundle). The CFD study has analyzed the maximum decay heat input Case (iii) scenario. In this scenario, a full-core discharge with a thirty day hold time is considered in which the 560 assemblies are transferred to the pool.
Converged temperature contour and velocity vector plots obtained from the FLUENT model are presented in Figures 5.8.7 and 5.8.8. The peak local temperature is conservatively estimated to be less than 334.7"K (143*F). This result conservatively ensures local subcooled conditions with a substantial margin of safety.
The peak cladding temperature is determined for the hottest cell location in the pool as obtained from the CFD model for the Oyster Creek pool, which includes the postulated dropped assembly scenario and 50% cell exit blockage. The maximum cladding-to-local water temperature difference (AT,) is calculated to be 24,1*F. Applying this calculated cladding A'l; to the peak local water temperature at the top of the active fuel length, a 167.1 F peak cladding temperature is determined. This is lower than approximately 239'F local boiling temperature on top of the racks.
Thus, nucleate boiling does not occur anywhere within the Oyster Creek pool.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Hollec international 5-17 Report HI-981983
5.9 Reactor Buildine HVAC An evaluation of the Reactor Building's HVAC System under the condition of maximum pool heat load (normal and full core discharge scenarios) is necessary to ensure that the humidity burden on the HVAC system is not exceeded from previous design basis. The maximum pool water temperatures, it is recalled from Table 5.8.1, for these two scenarios are 114.1*F and 124.6 F.
The pool design basis temperature (i.e.,125"F) for these conditions envelopes the transient temperature responses. Therefore, it is concluded that the previous design basis psychrometric burden on the HVAC system envelopes the peak loads computed in this licensing submittal. Any upgrading of the HVAC system is therefore not necessary.
5.10 References for Section 5
[5.5.1] Wang, Yu, " Heat Loss to the Ambient from Spent Fuel Pools: Correlation of Theory with Experknent", Holtec Report HI-90477, Rev. O, April 3, 1990.
[5.6.1] Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press,1967.
[5.6.2] Hinze, J.O., " Turbulence", McGraw Hill F? blishing Co., New York, NY, 1975.
[5.6.3] Launder, B.E., and Spalding, D.B., " Lectures in Mathematical Models of Turbulence", Academic Press, London,1972.
[5.6.4] "QA Documentation and Validation of the FLUENT Version 4.3 CFD Analysis Program", Holtec Report HI-951444.
[5.7.1] Rohsenow, N.M., and Hartnett, J.P., " Handbook of Heat Transfer",
McGraw Hill Book Company, New York,1973.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-18 Report HI-981983 j l
l
Table 5.1.1 PARTIAL LISTING OF RERACK APPLICATION USING SIMILAR METHODS OF THERMAL-HYDRAULIC ANALYSIS PLANT DOCKET NO.
Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254,50-265 Rancho Seco USNRC 50-312 Grand Gulf Unit 1 USNRC 50-416 Oyster Creek USNRC 50-219 Pilgrim USNRC 50-293 V.C. Summer USNRC 50-395 Diablo Canyon Units 1 and 2 USNRC 50-275, 50-455 Byron Units 1 and 2 USNRC 50-454, 50-455 Braidwood Units 1 and 2 USNRC 50-456,50-457 Vogtle Unit 2 USNRC 50-425 St. Lucie Unit 1 USNRC 50-425 Millstone Point Unit 1 USNRC 50-245 D.C. Cook Units 1 and 2 USNRC 50-315, 50-316 Indian Point Unit 2 USNRC 50-247 Three Mile Island Unit 1 USNRC 50-289 J. A. FitzPatrick USNRC 50-333 Shearon Harris Unit 2 USNRC 50-401 Hope Creek USNRC 50-354 Kuosheng Units 1 and 2 Taiwan Power Company Chin Shan Units 1 and 2 Taiwan Power Company i
SHADED TEXT CONTAINS PROPRIETARY INFORMATION liollec International 5-19 Report HI-981983 I
r i
- l. -
l I'
I l
l Table 5.1.1 (continued)
PARTIAL LISTING OF FUEL RERACK APPLICATIONS USING SIMILAR METHODS OF THERMAL-HYDRAULIC ANALYSIS PLANT DOCKET NO.
Ulchin Unit 2 Korea Electric Power Corporation Laguna Verde Units 1 and 2 Comision Federal de Electricidad Zion Station Units 1 and 2 USNRC 50-295,50-304 f Sequoyah USNRC 50-327, 50-328 l La Salle Unit One USNRC 50-373
! Duane Arnold USNRC 50-331 Fort Calhoun USNRC 50-285 Nine Mile Point Unit One USNRC 50-220 Beaver Valley Unit One USNRC 50-334 Limerick Unit 2 USNRC 50-353 l Ulchin Unit 1 Korea Electric Power Corporation Waterford 3 USNRC 50-382 l l
l l
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Internationa! 5-20 Report HI-981983 L
Table 5.3.1 OYSTER CREEK HISTORICAL AND PROJECTED FUEL DISCHARGE DATA Startup Date Shutdown Length of Reload Batch Cycle Date Operation Size (# of No. (Months) Bundles) j 1A 12/23/69 09/18/71 21.0 N/A IB 11/12/71 05/01/72 5.6 24 l 2 06/20/72 04/13/73 9.8 136 3 06/04/73 04/13/74 10.3 148 I 4 07/01/74 03/29/75 8.9 72 5 05/26/75 12/27/75 7.1 112 6 03/11/76 04/23/77 13.4 56 7 08/04/77 09/16/78 13.4 128 8 12/08/78 01/05/80 12.9 168 9 07/16/80 02/12/83 30.9 160 10 11/03/84 04/11/86 17.2 200 11 12/21/86 09/30/88 21.6 188 12 05/11/89 02/16/91 21.2 172 13 06/27/91 11/28/92 17.1 144 14 02/16/93 09/10/94 18.8 168 15 12/16/94 09/04/ % 20.7 172 16 10/23/ % 09/26/98 23.1 188 17 11/14/98 09/30/00 22.5 184 18 11/14/00 09/28/02 22.5 188 Note: Italics denote projected operation.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-21 Report HI-981983
1 l
1 l
l l
l l
Table 5.4.1
SUMMARY
OF KEY THERMAL-HYDRAULIC f DESIGN PARAMETERS l Reactor Power 1930 MWT Core Size 560 Assemblies .
Fuel Exposure Time, EFPD 2,053 days j
Fuel Transfer Rate 5.8 Assemblies /hr Radial Peaking Factor 1.50 l Total Axial Peaking Factor 2.01 Cell Opening 5.9305" i
Cell Length 169" ;
Bottom Plenum Height 7-1/2" j East Wall Downcomer Gap 3-15/16" West Walt Downcomer Gap 1-1/2" Fuel Pool Building Temperature 104 F SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-22 Report HI-981983
n-1 Table 5.4.2 i
OYSTER CREEK FUEL DESIGN DATA 1 i
Flow ChannelI.D. = 5.278"t Flow Channel O.D. = 5.438" i
Array Rods O.D. (d) Pitch (p) Hydraulic Fuel Type - [ inch] [ inch] Diameter (DJ
[ inch] l 7x7 GE I 0.570 0.738 0.647 7x7 SPC III-A 7x7 SPC III-E 7x7 SPC III-F 7x7 GE II 0.563 0.738 0.669 8x8 SPC 0.5015 0.642 0.545" V, & V3 8x8 GE 0.483 0.640 0.597 8x8 GE8B 8x8 GE9B 1
l Flow channel dimensions are applicable to all fuel types listed in this table.
" Bounding fuel assembly type hydraulic resistance.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-23 Report HI-981983
i l
Table 5.8.1 l
RESULTS OF BULK POOL TRANSIENT EVALUATION Coincident Coincident Coincident Peak Bulk Time (Hrs after Exchanger Heat Evaporation Discharge Temperature - Reactor Removal Heat Removal Scenario (*F) Shutdown) (Btu /hr) (Bru/hr) !
I 6
Case (i) 168.35 205 6.16x106 2.08x10 1 Partial Core Discharge,6 i Day Hold l
Case (ii) 114.10 187 8.40x10 6 0.06x106 Normal Discharge 6-day Hold '
Case (iii) 124.62 853 12.06x106 0.19x106 Full-Core Offload, 30-day Hold SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-24 Report HI-981983
Table 5.8.2 RESULTS OF LOSS-OF-COOLING I
Time-to-Boil (Hours)
Case Number (Without Makeup Water)
Case (i)' 7.75 Case (ii) 17.21 Case (iii) 10.49 i
I i
'l i
This case postulates loss of cooling from both the shell & tube heat exchanger and the augmented heat exchanger.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-25 Report H1-981983
+
)
I_
4 i 5_
\
i R I D
E m , (
4 3
3
\
G E D 0 R M N 89 A I A \S HR T CH ) U S/ E IS i l l
DA ( R P
F \
S S S O A8 E F
I 5 S R 8 A A 8T C N 1 A E O C I S R A
E N G E R RE C A G S H N C A E S HC G I
X R D E A I
I E T =
C R A S O EH I D
C E N L N N L A O O O I
D O D T P T N U R A I I
S L A
P PM R E
U = O U
) F P T C C 0 E A T E NO E N R E
(
P E S L
C K Y E C E R
F C O
S R D S E N T E S Y
O 1
4 5
L E
4 R 82 5 $h5 dE U G
I F
c .
t 2_
4 S_
\
R _
l
\
D 4 E m V 3 3
G E 0 R M 8 9
A I N _
S/ i l I S [
DA i R P F ( N S .
A8 E F _
S 5
S 0 A O 6T C I 5A RR O AE I NG R _
E NA A C H N SC E _
X C EE S
G _
R AT E AE G HH R C A S LO H I
O = C DP S I
EL D R EU N _
O F W L _
C O O _
D D O L E T i
P L TN I H
U E S L FM R E
C = O U
) U T F C A C 0 D A T EN E N
( A R E P
F S C
_ OE N $ R D O $ E
_ N T
_ E S Y
O 2
_ 4 L
5 A
E R
d2 Ee E y5 dE U G
_ I F
r .
l
\
l i .
I I
l EVAPORA110NIMATLOSS
(
SPENTFUEL POOL 1
1 HEATEXCHANGER Tg P
-y COOLANT t, M(
FIGURE 5.5.1: SPENT FUEL POOL COOLING MODEL
0 0
8 E
0 G 0 R 7 A H _
- C S
I D _
E 0 R _
0 O _
6 C _
L A
I T
& R l )
s A le r P h
S 0H R 0 (
1 5 n O w F d o T n d O a t u L p h P m S E u r R
P )j r 0
0 t o
U 1
( e 4 ca T g A n e oa R R ira:h r E e P nx f t
M eE ct A E S a 0 e T e 0 T
)
3 m i H
( i T N ee sb E
I a u S CT N A
R 0 T 0 L 2 O O
P 1 K L
U B
0 :
0 1 1
8 5
E R
U G
I
- F
- . 0 0 0 0 0 6 4 2 0 1 1 1 1
!3EE F
0 0
8 E
- 0 G 0
7 R
A .
H _
C _
S N
I D
0 L I
0 A 6 M _
d R t
e O n
N e
A d
g n
m u % 0 0
5
)
H
(
s r
n w
R O
F T
O _
a o L _
dt P _
p u E P
1
(
m u
o ,
N 0 r h
S 0 to 4 ca R
T A
R E
U e
i) r a
' R P nge r
r M e E en t T
Sc ha f
0 A T
)
i cx 0 e N i( 3 m E E i I S
et T s a N ae A CH R T
0 L 0
UC 0
B S 0
1 2
8 5
E R
U G
I
- 0 F 0 0 0 1 0 9 1 1
% e2E&E F
0 0
0 1
d t
e n
e 0 E m 2 0 R g 8 O u C A
d n
/ L L .
a U .
p F .
m R _
u ) O _
P s r
F _
1 H T _
(
( O _
L .
io\
r 0 n 0 w P _
a r 6 o E n )e eg dt R cn u U h T Sa h S A
) _
i i
r r o R _
iet E
( t c E .
a P _
s a e M ae R E CH r T e T NIO t
0 f 0 A ER 4 e S A I
m N i
T N A E RC S T
0 UI S 0 BD 2 :
3 8
5 E
R U
G I
F
. - 0 0 0 0 0 2 1 0 9
_ 1 1 1
$ $3&E -i
0 0
8 L I
A T
R A
P
_ )
~
i i
0 (
0 E 7 S A
C D
N A
_ 0 0
)
i
(
6 E S
A C
N 0 0
)
(
s 5 n A r O H F w LO R
D do t T hu A E
S H r
0 o Y 0
4 ca CA t
e E Rr D e T NS N
t f
A EO I 0
0 e D R 3 imNA EN T PE EC N 0 DS ED MA TOL I
0 b 2 L OFF OO P E i
L R E
U O 0 FC 0 :
1 4 8
5 E
R U
G I
. ~ - ~ 0 F 9
- 7 6 5 4 3 2 1 0
- E 3m 6.m v i x8 o
0 0
0 1
0 E 0 R 9 O C
A L L
U 0 F 0 )
8 i i
i
(
/ E S
A 0 C 0 R 7
)
s O r F H(
D A
0 n O 0 w L 6 od T t
u A h E S H r Y 0 o A 0 c 5 a t
C E
e D R T r
e N 0 f t
E 0 A D 4 e N m E i
P T EO I DR 0 E A 0
3 MN I
0 OD A P O 0
- 2 LL EF UF FO:
5 0 8 0
1 5 E
R U
G
. ~ - 0 FI 4 2 0 8 6 4 2 0 1 1 1
$3m8=Esyx38 1
0 6
1 0
4 S h T 1
O L
P E
)
i o i i i( r en ac/
a
% 0 2
MI T.
g s
s e 1 v
H CS
/ )
H s
r E T
P D
/ 0 g 0 in 1 lo
(
O L
O
/ % N \ C d
e o P G
IN
/ 0 F r
o c L O
O hX x 8 f C o D s E s
)
i io o C i
( r L R a
e s e n r e O F
ac/
%N t
f CS 0 6 A F
O
/ e im T
S S
O N L T
g/h S
)
i i o 0 O
( r 4 P e a s e n/ 6 a c 8 CS 5 N E R
W " " "
0 2
0 U
G I
F
" """""2 "2 2 0 2
g1a
8 9 2 c .
9 3 n 1 4 I t
5 t n n 2
g u ue e
l AuF F l
4s .
eWIllI llIlj1' 4q-msEm4 3i(ijSewy.. d ,... .-o,ih,ge [/^
.\
- ?.,
d
- g%i'. "'
1@I _
.m)?p a Ne,g e -
.v
- t .
sm; Q x>i; %M,Yy hgWmx
..NM;m.W3G2yky.yC V n.Mr 3a j p A .. p J / ..
, mY:.* ,.Q. slshrf. u <.mIvmnd.nkg ae. %
.*aY
- My.W=' t~ _
g r we
- t. dp/ .A
@T+,4p$g&nd ,t:p)dxBp+Fon.v,9 g
[# < _
t 7 F gr p j :
t
/4w _y.n. WkM. ~ _
e
%5 - q' .F. ~ -
mri hs-
.. s
? ;t e~
.r2 r- -
- /gn<%.f-.M
- JuW.qtb.'J.A.3 s. ,.r mW5-am 3' _
.r tl' .
~ .
. . 'tQMJ2 v; . b . _ ./ }
. I dmy.%:
2 s
.JyNy;m.eM1.MgM .g%y p w W@ p ? g..eN' Mi _
I V
W'3w -@.ky+m . ;
\
r u
o
- fgAy9MgQ[i3c.Wg~k.ws.3Mym dp" i fN.
d _ .
. t n
3i.adyW;4Au a t.et t
p J &fgMg?c,w o i
.~.9 7_ w s
- g M.sNM'g$ Wy o
_4m/;M d agy,k.ukg f
.g d sf M;
5 ;;m,V4$kpa.-
nN,.A .A;' .- y.p;,Lz**_-C .
f.
Wi ,@t tc J- .O.1dWw!-
- ,b - b qg. w -
- ey, t e
r
- 5v t F4 es*m%wd $wA'hv F u L
4$d wd+:f%1 n .Ip* 7" *el [iM yrh;:.
t _. t a
_7. V' dKpngn.Dt g;z.
, E. ii
~
- ~
c
_. e r p
e.
s
-@f;FiMg gv+a 3 <J~ 7 Mg4,~ 'Q+'"
4 ., .Md.'"
m i.'
_ f. _ e
.7 ;'M- Mnf y f kYu . ,s;:.
t s
'e z" T
.' 2 9.A s.
.r l . t~ _ l o 2
,E;?e+M.p&g .s mfM3Vi" 3fg7 t _ 0
~, qQ ,. .
$ ~
o +
P
,N 1 n. .._ g$nWMp. No" E
.s}qi y- ."
i,s e
q e-3w t
- _ l 3
e 7 r
.' _ u 0 4
F 4
mxM _.
k 3
_ e =
~
.e s. .c
- e r n
- ,pc;m . A.;~ -
C i y
.M n!
M L'1wa4,gN
'*hgn;: 7..~2.3
,{ r e
1 t
s) 2 tr . .( .+ :
g yK 0
- cc.3:.o! kt y j(& gys,5N
.A-
! mrEE -y,_ . .v 0
<? :: } +n lllyll O( +E l
e7 r
7 8 t u43 5 ra e 3 e
r p =
i u emxa g
F TM 2 2 2 2 2 2 2 2 2 2 22 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 0 0000 00 00
+ 4 + + ++ + + +
EEEEEEEEEEEEEEEEEEEEEEEEEEEEEEE 000+++000 + + + +000+++++00 0 00 0 0 0 0 0 0 0 0 4 + + + + +++++ x 5 4 3 2 1 0 9 8 7 6 6 5 4 3 2 1 0 9 8 7 6 6 5 4 3 2 1 0 9 8 7 3 3 3 3 3 3 2 2 2 2 2 2 2 2 2 2 2 1 1 1 1 1 1 1 1 1 1 1 0 0 0 3 3 3 3 3 3 3 3 3 3 33 3 3 3 3 3 3 3 33 3 3 3 3 33 3 3 3 3 Y
^ g h,.%"~
, %,N ,m,#^
8 _
9 2 c . -
9 3 n 1 4 I t
5 tn n 2 e e g ulu AuF F l
y\. ' ; I.
l
[fI ,\ , , i ' ' ' '
' ? P f t' f l11 t
4n4 s
'(
l ,=
C' N yyyN% '
.&,T. _
~ [
1 _
_ Z~qsss x \ \ \ i ' '
v-
'~2~ ~ ~ ' '
~
[ , _ - -
Z~2~~ , , _ _
- r
'~,_ -
. - ~ ' ' , ,. _ _ _ - -
d l
e i
T F
- y
~ =-~ -
' ' - t
- i c
s l
o
=-~~' -
e
~ - V r
a l
=
-~- c
, _ _ __ _ - - o
. r L
- l o
_--~ -
- ~ o 4
, , , ~ ~ 0-T P
2 -
l E
. - e 5
_ - ~ , ,
, s -t u 9 s ~. _ F 3 1-.~- -
k 7 e
' , , s , , s s ss e r
=
y""Q,"*'f n
_ ^ H' L
C) Si
. j t g w n r /M
- i"fsf%
/ ;
* " ./ gUjg g g g s l g g w : eM
. jf'l' j'f.
uum%w%%w+
' ii \
g \
l, \ \ N \s w
' . e t
s( 1 y s0r
- -\
w- OoE
' \ =
3 l# ' l l m-
[ ' , , ,
l
' \'
7]- l
- lllI : t c2 8 e8 8 V1 5 y2t ei r c =
u o x gl e a i
F VM 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 22 2 2 2 2 2 2 2 2 2 2 3 4 0- 0 - -000 - - -000 - - - 0 -00 - 0 00000 0 0 00 0 00 0 00 0 0 - - - - - - - x EEEEEEEEEEEEEEEEEEEEEEEEEEEEEE 8 1 3 6 8 1 3 6 8 1 3 6 8 1 3 6 2 7 2 7 2 7 2 7 2 7 2 7 4 9 1 1 0 9 8 8 7 6 5 5 4 3 2 2 1 0 8 0 3 5 8 0 3 5 8 0 3 5 2 3 2 2 21 1 1 1 1 1 1 1 1 1 1 1 1 9 9 8 7 6 6 5 4 3 3 2 1 8 7 .
3j r i >;, i l
.i '
n.
M '- k
6.0 STRUCTURAUSEISMIC CONSIDERATIONS L 6.1 Introduction This section considers the structural adequacy of the new maximum density spent fuel racks l under all loadings postulated for normal, seismic, and accident conditions at Oyster Creek Nuclear Generating Station (OCNGS). The existing spent fuel storage racks are also examined for stability during the installation process. The analyzed storage rack configuration with the new racks in place is shown in Figure 1-1.
1 The analyses undertaken to confirm the structural integrity of the racks are performed in compliance with the USNRC Standard Review Plan (6.1.1] and the OT Position Paper [6.1.2].
l l For each of the analyses, an abstract of the methodology, modeling assumptions, key results, and i summary of parametric evaluations are presented. Delineation of the relevant criteria is discussed in the text associated with each analysis.
6.2 Overview of Rack Structural Analysis Method The response of a free-standing rack module to seismic inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting in impacts I and friction effects. Some of the unique attributes of the rack dynamic behavior include a large fraction of the total structural mass in a confimed rattling motion, friction support of rack pedestals against lateral motion, and large fluid coupling effects due to deep submergence and independent motion of closely spaced adjacent structures.
1 I
Linear methods, such as modal analysis and response spectrum techniques, 'cannot accurately L simulate the structural response of such a highly nonlinear structure to seismic excitation. An l accurate simulation is obtained only by direct integration of the nonlinear equations of motion with the three pool slab acceleration time histories applied as the forcing functions acting simultaneously.
i SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-1 Report Hi-981983
)
i 1
L l .Whole Pool Multi-Rack (WPMR) analysis'is the vehicle utilized in this project to simulate the l
dynamic' behavior of the complex storage rack structures. The following sections provide the basis for this selection and discussion on the development of the method.
6.2.1 Background of Analysis Method Reliable assessment of the stress field and kinematic behavior of th6 rack modules calls for a L ~ conservative dynamic model incorporating all key attributes of the actual structure. This means l
that the model must feature the ability to execute the concurrent motion forms compatible with the free-standing installation of the modules.
The model must possess the capability to effect momentum transfers which occur due to rattling l of fuel assemblies inside storage cells and the capability to simulate lift-off and subsequent impact of support pedestals with the pool liner (or bearing pad). The contribution of the water mass in the interstitial spaces around the rack modules and within the storage cells must be l modeled in an accurate manner since erring in quantification of fluid coupling on either side of the actual value is no guarantee of conservatism.
1 The Coulomb friction coefficient at the pedestal to pool liner (or bearing pad) interface may lie in l l a rather wide range and a conservative value of friction cannot be prescribed a priori. In fact, a perusal of results of rack dynamic analyses from numerous dockets (Table 6-1) indicates that an upper ' bound value of the coefficient of friction often maximizes the computed rack displacements as well as the equivalent elastostatic stresses.
In short, there are a large number of parameters with potential influence on the rack kinematics.
The comprehensive structural evaluation must deal with all of these without sacrificing conservatism.
The three-dimensional single rack dynamic model mtroduced by Holtec International in the i Enrico Fermi Unit 2 rack project (ca 1980) and used in some 50 rerack projects since that time (Table 6-1) addresses most of the above mentioned array of parameters. The details of this l
I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International _ 62 Report HI-981983
L methodology are also published in the permanent literature [6.2.1). Despite the versatility of the 3-D seis'mic model, the accuracy of the single rack simulations has been suspect due to one key element; namely, hydrodynamic participation of water around the racks. During dynamic rack motion, hydraulic energy is either drawn from or added to the moving rack, modifying its submerged motion in a significant manner. Therefore, the dynamics of one rack affects the motion of all others in the pool.
A dynamic simulation that treats only one rack, or a small grouping of racks, is intrinsically inadequate to predict the motion of rack modules with any quantifiable level of accuracy. Three-dimensional Whole Pool Multi-Rack analyses carried out on several previous plants demonstrate that single rack simulations under predict rack displacements [6.2.2].
Briefly, the 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the array of variables as follows:
Interface Coefficient of Friction: Parametric runs are made with upper bound and lower bound values of the coefficient of friction. The limiting values are based on experimental data, which are bounded by the values 0.2 and 0.8. Simulations are also performed with the array of pedestals having randomly chosen coefficients of friction in a Gaussian distribution with a mean of 0.5 and lower and upper limits of 0.2 and 0.8, respectively. In the fuel rack simulations, the Coulomb friction interface between rack support pedestal and liner is simulated by piecewise linear (friction) elements. These elements function only when the pedestal is physically in contact with the pool liner.
_ Rack Beam Behavior: Rack elasticity, relative to the rack base, is included in the model by introducing linear springs to represent the elastic bending action, twisting, and extensions.
i l
l t
SHADED TEXT CONTAINS PROPRIETARY INFORMATION
- Holtec International 6-3 Report HI 981983 ,
1 1
1 l
1 i
Imnact Phenomena: Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface. These interface gaps are modeled using nonlinear spring elements. The term
" nonlinear spring" is a generic term used to denote the mathematical representation of the
]'
condition where a restoring force is not linearly proportional to displacement.
Fuel Loadinn Scenario: The fuel assemblies are conservatively assumed to rattle in unison, which exaggerates the inertia effect due to fuel impacts with the cell walls.
n Holtec International extended Fritz's classical two-body fluid coupling model to Fluid Coupling; multiple bodies and utilized it to perform the first two-dimensional multi-rack analysis (Diablo Canyon, ca 1987). Subsequently, laboratory experiments were conducted to validate the multi- !
-rack fluid coupling theory. This technology was incorporated in the computer code DYNARACK (a.k.a. MR216) [6.2.4] that handles simultaneous simulation of all racks in the pool as a Whole Pool Multi-Rack 3-D analysis. This development was first utilized for l l
Chinshan, Harris, Oyster Creek [6.2.1, 6.2.3] and, subsequently, for numerous other rerack 4 projects. The WPMR analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses, and also serve to improve predictions of rack kinematics.
For closely spaced racks, demonstration of kinematic compliance is verified by including all modules in one comprehensive simulation using a WPMR model. In WPMR analysis, all rack modules are modeled simultaneously and the coupling effect due to this multi-body motion is included in the analysis. Due to the superiority of this technique in predicting the dynamic )
behavior of closely spaced submerged storage racks, the Whole Pool Multi Rack analysis methodology is used for this project.
I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-4 Report HI-981983
p I
l 6.3 Descriotion of Racks The implementation of the storage capacity increase is intended to be completed in one phase (campaign). Figure 6-1 shows the fully implemented rack configuration and assigns identification numbers to the racks for later analysis. The Cartesian coordinate system utilized within the rack dynamic modelis defined as follows:
X- =
Horizontal axis parallel to plant West Y =
Horizontal axis parallel to pl. nit South Z =
Vertical axis upward from the rack base The materials utilized in fabrication of the rack components are identified in Table 6-2.
6.4 Synthetic Time '4istories The synthetic time-h. xies in three orthogonal directions (N-S, E-W, and vertical) were generated in accordan( with the provisions of SRP 3.7.1 [6.4.1]. The SSE time histories were developed in accordance with the requirements of SRP 3.7.1 [6.4.1] by EQE International
[6.4.3]. In order to prepare an acceptable set of OBE acceleration time histories, Holtec l International's proprietary code GENEQ [6.4.2] was utilized.
l l
A preferred criterion for the synthetic time-histories in SRP 3.7.1 calls for both the response spectrum and the power spectral density corresponding to the generated acceleration time history to envelope their target (design basis) counterparts with only finite enveloping infractions. The OBE time histories for the pool have been generated to satisfy this preferred criterion. The seismic files also satisfy the requirements of statistical independence mandated by SRP 3.7.2
[6.4.4].
Figures 6-2 through 6-4 and 6-5 through 6-7 provide plots of the accelerograms, which were generated over a 20-second duration, for the OBE and SSE events, respectively.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-5 Report HI-981983
i I'
6.5 WPMR Methodolony Recognizing that the analysis work effort must deal with both stress and displacement criteria, the sequence of model development and analysis steps that are undertaken are summarized in the i
l following:
- a. Prepare 3-D dynamic models suitable for a time-history analysis of the new maximum density racks. These models include the assemblage of all rac) modules in the pool. Include all fluid coupling interactions and mechanical coupling appropriate to perform an accurate non-linear simulation. This 3-D simulation is referred to as a Whole Pool Multi-Rack model.
l
- b. Perform 3-D dynamic analyses on various physical conditions (such as coefficient of friction and extent of cells containing fuel assemblies). Archive appropriate j displacement and load outputs from the dynamic model for post-processing.
- c. Perform stress analysis of high stress areas for the limiting case of all the rack dynamic analyses. Demonstrate compliance with ASME Code Section III, Subsection NF limits on stress and displacement.
6.5.1 Model Details for Soent Fuel Racks The dynamic' modeling of the rack structure is prepared with special consideration of all nonlinearities and parametric variations. Particulars of modeling details and assumptions for the Whole Pool Multi-Rack analysis of racks are given in the following:
6.5.1.1 Assumotions
- a. The fuel rack structure motion is captured by modeling the rack as a 12 degree-of-freedom structure. Movement of the rack cross-section at any height is described by six degrees-of-freedom at the rack base and six degrees-of-freedom at the rack top. In this manner, the response of the module, relative to the baseplate, is captured in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom and simulate rack stiffness.
- b. Rattling fuel assemblies within the rack are modeled by five lumped masses located at H, %H, %H, %H, and at the rack base (H is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed equal SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-6 Report HI-981983
L l-l l
to rack vertical motion at the baseplate level. The centroid of each fuel assembly
. mass can be located off-center, relative to the rack structure centroid at that level, to simulate a partially loaded rack. For the dynamic simulations, the dry weight of a BWR fuel assembly is set equal to 690 lb.
- c. Seismic motion of a fuel rack is characterized by random rattling of fuel assemblies in their individual storage locations. All fuel assemblies are assumed
. to move in-phase within a rack. This exaggerates computed dynamic loading on the rack structure and, therefore, yields conservative results.
- d. Fluid coupling between rack and fuel assemblies, and between rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy. Inclusion l of these effects uses the methods of [6.5.2,6.5.3] for rack / assembly coupling and for rack-to-rack coupling.
- e. Fluid damping and form drag are conservatively neglected.
- f. Sloshing is found to be negligible at the top of the rack and is, therefore, neglected l in the analysis of the rack.
- g. Potential impacts between the cell walls of the new racks and the contained fuel assemblies are accounted for by appropriate compression-only gap elements between masses. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at the top and bottom of the rack in two horizontal l directions. Bottom gap elements are located at the baseplate elevation. The initial gaps reflect the presence of baseplate extensions, and the rack stiffnesses are l chosen to simulate local structural detail.
l l h. The rack model is supported, at the base, by four pedestals, which are modeled as
! nonlinear compression-only gap elements in the vertical direction. In the i
horizontal direction, the pedestals act as " rigid links." Each pedestal support is linked to the pool liner (or bearing pad) by two friction springs.
- i. . Rattling of fuel assemblies inside the storage locations causes the gap between l fuel assemblies and cell wall to change from a maximum of twice the nominal gap l to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap in order to provide a conservative measure of fluid resistance to gap closure.
I 1
I i
l I
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-7 Report HI-981983 L
I l l
6.5.1.2 Element Details Figure 6-8 shows a schematic of the dynamic model of a single rack. The schematic depicts many of the characteristics of the model including all of the degrees-of-freedom and some of the spring restraint elements.
i l Table 6-3 provides a complete listing of each of the 22 degrees-of-freedom for a rack model. Six translational and six rotational degrees-of-freedom (three of each type at each end) describe the
{
motion of the rack structure. Rattling fuel mass motions (shown at nodes 1,2*,3*,4', and 5* in I Figure 6-8) are described by ten horizontal translational degrees-of-freedom (two at each of the l five fuel masses). The vertical fuel mass motion is assumed (and modeled) to be the same as that j of the rack baseplate.
l Figure 6-9 depicts the fuel to rack impact springs (used to develop potential impact loads between the fuel assembly mass and rack cell inner walls) in isometric view. Only one of the five fuel masses is shown in this figure. Four compression only springs, acting in the horizontal direction, are provided at each fuel mass.
Figure 6-10 provides a 2-D schematic elevation of the storage rack model, discussed in more detail in Subsection 6.5.3. This view shows the vertical location of the five storage masses and l some of the support pedestal spring members.
l Figure 6-1I shows the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending plane a shear and bending spring simulate elastic effects [6.5.4].
l Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model.
Figure 6-12 depicts a single rack module with its surrounding impact springs (used to develop potential impact loads between racks or between rack and wall).
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-8 Report HI-981983
m 6.5.2 Eluid Coupline Effect In its simplest form, the so-called " fluid coupling effect" (6.5.2, 6.5.3] can be explaind by considering the proximate motion of two bodies under water If one body (mass mi) vibrates adjacent to a second body (mass m2), and both bodies are submerged in frictionless fluid, then
- Newton's equations of motion for the two bodies are:
(m, + M )f +i M Ii2 ii = applied 2 forceson massmi + O(Xi ')
M2i I +(m2 t + M )I22= applied 2 forces on mass m2 + O(X2 *)
S iand"X denote 2 absolute accelerations of masses mi and m2, respectively, and the notation 2
O(X ) denotes nonlinear terms.
M ii, M , i2M , and 2i M22 are fluid coupling coefficients, which depnd on body shape, relative disposition, etc. Fritz [6.5.3] gives data for My for various tx>dy shapes and arrangements. The fluid adds mass to the body (M ii ot mass mi), and an inertial force proportional to acceleration of the adjacent body (mass m2). Thus, acceleration of one body affects the force field on another.
This force field is a function of inter-body gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a storage location encounters this effect. For example, fluid coupling behavior will be experienced between nodes 2 and 2* in Figure 6-8. The rack analysis also contains inertial fluid coupling terms, which model the effect of fluid in the gaps between adjacent racks.-
, Terms modeling the effects of fluid flowing between adjacent racks in a single rack analysis suffer from the inaccuracies described earlier. These terms are usually computed assuming that all racks adjacent to the rack being analyzed are vibrating in-phase or 180* out of phase. The WPMR-analyses do not require any assumptions with regard to phase. Rack-to-rack gap elements have initial gaps set equal to 100% of the physical gap between the racks or between the outermost racks and the adjacent pool walls.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION l Holtec International 6-9 Report Hi-981983
6.5.2.1 Multi-Body Fluid Couplina Phenomena l During the seismic event, all racks in the pool are subject te the input excitation simultaneously, i The motion of each free-standing module would be autonomous and independent of others as long as they did not impact each other and no water were present in the pool. While the scenario I
)
i ofinter-rack impact is not a common occurrence and depends on rack spacing, the effect of water
. the so-called fluid coupling effect - is a universal factor. As noted in Ref. [6.5.2,6.5.4], the fluid forces can reach rather large values in closely spaced rack geometries. It is, therefore, essential that the contribution of the fluid forces be included in a comprehensive manner. This is possible only if all racks in'the pool are allowed to execute 3-D motion in the mathematical j
model. For this reason, single rack or even multi-rack models involving only a portion of the racks in the pool, are inherently inaccurate. The Whole Pool Multi-Rack model removes this
{
intrinsic limitation of the rack dynamic models by simulating the 3-D motion of all modules simultaneously. The fluid coupling effect, therefore, encompasses interaction between every set of racks in the pool, i.e., the motion of one rack produces fluid forces on all other racks and on the pool walls. Stated more formally, both near-field and far-field fluid coupling effects are included in the analysis.
The derivation of the fluid coupling matrix [6.5.5] relies on the classical inviscid fluid mechanics principles, namely the principle of continuity and Kelvin's recirculation theorem. While the derivation of the fluid coupling matrix is based on no artificial construct, it has been nevertheless verified by an extensive set of shake table experiments [6.5.5].
6.5.3 Stiffness Element Details Table 6-4 lists all of the spring elements used in the 3-D WPMR model for Oyster Creek. Three element types are used in the rack models. Type 1 elements are linear elastic elements used to represent the beam-like behavior of the integrated rack cell matrix. Type 2 elements are the piece-wise linbar friction springs used to develop the appropriate forces between the rack pedestals and the bearing pads. Type 3 elements are nonlinear gap elements, which model gap SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-10 Report HI 981983 i
r closures and subsequent impact loadings (e.g., between fuel assemblies and the storage cell walls). '
If the simulation model is restricted to two dimensions (one horizontal motion plus one vertical motion, for example), for the purposes of model clarification only, then Figure 6-10 describes the configuration. This simpler model is used to elaborate on the various stiffness elements.
Type 3 gap elements modeling impacts between fuel assemblies and racks have local stiffness Ki in Figure 6-10. In Table 6-4, for example, type 3 gap elements 5 through 8 are connected to the rattling fuel mass at the rack top. Support pedestal spring rates Ks are modeled by type 3 gap elements I through 4, as listed in Table 6-4. Local compliance of the concrete floor is included in Ks. The type 2 friction elements listed in Table 6-4 are designated in Figure 6-10 by the label Kr. The spring elements depicted in Figure 6-11 represent type 1 elements.
Friction at support / liner interface is modeled by the piecewise linear friction springs with suitably large stiffness Kr up to the limiting lateral load pN, where N is the current compression load at the interface between support and liner. At every time-step during transient analysis, the current value of N (either zero if the pedestal has lifted off the liner, or a compressive finite value) is computed.
The gap element Ks, modeling the effective compression stiffness of the structure in the vicinity of the support, includes stiffness of the pedestal, local stiffness of the underlying pool slab, and local stiffness of the rack cellular structure above the pedestal. I' The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses
~ incorporate 3-D motions and include all stiffness elements listed in Table 6-4.
i I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec international ' 6-11 Report HI 981983
6.5.4 Coefficients of Friction !
To eliminate the last significant element of uncertainty in rack dynamic analyses, multiple single rack simulations are performed with different friction coefficients ascribed to the support pedestal / bearing pad interface. These friction coefficients are chosen consistent with the two l
bounding extremes from Rabinowicz's data [6.5.1]. Whole Pool Multi-Rack simulations are also performed with intermediate friction coefficients, which are developed by a random number generator with Gaussian normal distribution characteristics. The assigned values are then held constant during the entire. simulation in order to obtain reproducible results.' Thus, in this manner, the WPMR analysis results are brought closer to the realistic structural conditions.
According to Rabinowicz [6.5.1], results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of to be 0.503 with standard deviation of 0.125.
Upper and lower bounds (based on twice standard deviation) are 0.753 and 0.253, respectively. l a
Analyses are therefore performed for coefficient of friction values of 0.2 (lower limit) and for 0.8 )
i (upper limit), and for random friction values clustered about a mean of 0.5. i 6.5.5 Governine Eauations of Motion Using the structural model discussed in the foregoing, equations of motion corresponding to each l i
degree-of freedom are obtained using Lagrange's Formulation [6.5.4]. The system kinetic energy 1 includes contributions from solid structures and from trapped and surrounding fluid. The final 1 i
system of equations obtained have the matrix form:
I
[M] = [Q) + [G]
. dI .
l l
' It is noted that MR216 has the capability to change the coefficient of friction at any pedestal at each !
instant of contact based on a random reading of the computer clock cycle. However, exercising this l option would yield results that could not be reproduced. Therefore, the random choice of l coefficients is made only or.ce per run. l 1
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-12 Report HI-981983 I
i
where:
[M] -
total mass matrix (including structural and fluid mass contributions). The size of this matrix will be 22n x22n for a
.WPMR analysis (n = number of racks in the model).
I q -
the nodal displacement vector relative ta the pool slab displacement (the term with q indicates the second derivative with respect to time, i.e., acceleration)
[G] -
a vector dependent on the given ground acceleration
[Q] -
a vector dependeni on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom The above column vectors have length 22xn. The equations can be rewritten as follows:
I l
"OfI'[9]+0fJ'[G]
This equation set is mass uncoupled, displacement coupled at each instant in time. The numerical solution uses a central difference scheme built into the proprietary computer program MR216 [6.2.4].
6.6 Structural Evaluation of Spent Fuel Rack Design 6.6.1 Kinematic and Stress Acceptance Criteria There are two sets of criteria to be satisfied by the rack modules:
l
- a. Kinematic Criteria l Per Reference [6.1.1], in order to be qualified as a physically stable structure it is l necessary to demonstrate that an isolated rack in water exhibits safety factors of 1.5 and 1.1 against overturning during the OBE and SSE events, respectively.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-13 Report HI-981983 t.
l I
l
) b. Stress Limit Criteria Stress limits must not be exceeded under the postulated load combinations provided herein.
6.6.2 Stress Limit Evaluations The stress limits presented below apply to the rack structure and are derived from the ASME Code,Section III, Subsection NF [6.6.1], as amended by the requirements of paragraphs C.2, C.3 and C.4 of Regulatory Guide 1.124 [6.6.5]. Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code, 5ccion 11, Part D
[6.6.2], and are listed in Table 6-2.
(i) Nomial and Unset Conditions (Level A or Level B) i
- a. Allowable stress in tension on a net section is:
Fi = 0.6 S y where Sy is the yield stress at temperature, and F is equivalent to primary i
membrane stress,
- b. Allowable stress in shear on a net section is:
F, = 0.4 S y
- c. Allowable stress in compression on a net section ki '
Fa = Sy 0.47 -
r 444 rs kl/r for the main rack body is based on the full height and cross section of the honeycomb region and does not exceed 120.
1 =' unsupported length of component k= length coefficient which gives influence of boundary conditions. The following values are appropriate for the aescribed end conditions:
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-14 Report HI-981983 l
=
1 (simple support both ends)
=
2 (cantilever beam)
=
K (clamped at both ends) r= radius of gyration ofcomponent
- d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is:
F3 = 0.60 S y (equivalent to primary bending)
- f. Combined bending and compcession on a net section satisfies:
Q C ,fs, , C.,,f,, ,
F. D, Fs, D, Fs, where:
f, =
Direct compressive stress in the section
=
fox Maximum bending stress along x-axis
=
fsy Maximum bending stress along y-axis
=
Cmx 0.85 Cmy =- 0.85 Dx
= 1 - (f /F,x)
Dy =
1 - (f,/ Fey)
= 2 F,x,,y (B E)/[2.15 (ki/r)2x,y)
E =
Young's Modulus and subscripts x,y reflect the paiticular bending plane.
- f. Combined flexure and compression (or tension) on a net section:
la + bt + lby < j,o
- 0. 6 S, Fe, Fs,
- The above requirements are to be met for both direct tension or compression.
- g. Welds Allowable maximum shear stress on the net section of a weld is given by:
, F - 0.3 S, SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-15 Report HI-981983
e 1 i
where So is the weld material ultimate strength at temperature. For fillet weld legs in contact with base metal, the shear stress on the gross section is limited to 0.4Sy ,
yhere Sy is the base material yield strength at temperature.
l (ii)- Level D Service Limits j Section F-1334 (ASME Section III, Appendix F) [6.6.4] states that the limits for the l Level D condition are the smaller of 2 or 1.167S /Sy times the corresponding limits for l the Level A condition if Su is greater than 1.2S y , or 1.4 times the Level A limits if S is i l less than or equal to 1.2Sy , except for the requirements specifically listed below. Su and S y are the ultimate strength and the yield strength at the specified rack design temperature. j Examination of material properties for 304L stainless demonstrates that 1.2 times the I
- yield strength is less than the ultimate strength.
Exceptions to the above general multiplier are the following:
a) Stresses in shear shall not exceed the lesser of 0.72Syor 0.42So. In the case of the austenitic stainless material used here,0.72Sy governs.
b) Axial compression loads shall be limited to 2/3 of the calculated buckling load. j l
c) For combined axial compression and bending, the equations for Level A conditions shall apply except tM.
F = 0.667 x Buckling Load / Gross Section Area and the terms F,x and F,y may be increased by a factor of 1.65.
d) For welds, the Level D allowable maximum weld steess is not specified in Appendix ;
F of the ASME Code. An appropriate limit for weld throat stress is conservatively set i here as:
F. = 0.4 S, In addition, certain areas of the rack have been evaluated using some of the plate and shell rules presented in ASME NF-3200.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-16 Report HI-981983
=
6.6.3 vimensionless Stress Factors For convenience, the stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value. The limiting value of each stress factor is 1.0 based on an evaluation which uses the allowable strength appropriate to Level A or Level D loading as discussed above.
I The stress factors reported are:
R= i Ratio of direct tensile or compressive stress on a net section to its allowable value !
(note pedestals only resist compression)
R= 2 Ratio of gross shear on a net section in the x-direction to its allowable value R= 3 Ratio of maximun i:-axis bending stress to its allowaHe value for the section R= 4 Ratio of maximum y-axis bending stress to its allowable value for the section R= 5 Combined flexure and compressive factor (as defined in the foregoing)
R= 6 Combined flexure and tension (or compression) factor (as defined in the foregoing)
R= 7 Ratio of gross shear on a net section in the y-direction to its allowable value 6.6.4 Loads and Loading Combinations for Soent Fuel Racks The applicable loads and their combinations, which must be considered in the seismic analysis of rack modules, are excerpted from References (6.1.2) and [6.6.3].
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-17 Report HI-981983
5 j
1 i
The load combinations considered are identified below:
Loading Combination Service Level l D+L Level A D+L+To l D + L + To + E D + L + T, + E Level B -
D + L + To + Pr D + L + T. + E' Level D 1
D + L + To + F4 The functional capability of the fuel racks j must be demonstrated.
)
D= Dead weight-induced loads (including fuel assembly weight) i 4
L = Live Load (not applicable for the fuel rack, since there are no moving objects in the rack loadl 'ath) l Pr.= U Pward force on the racks caused by postulated stuck fuel assembly F4= Impact force from accidental drop of the heaviest load from the maximum possible height.
E =
Operating Basis Eannquake (OBE)
E' = Safe Shutdown Earthquake (SSE)
To = Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition)
T. = Differential temperature induced loads (the highest temperature associated j with the postulated abnormal design conditions)
T, and To produce local thermal stresses. The worst thermal stress field in a fuel rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated l rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing maximum possible temperature difference SIIADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-18 Report HI-981983
between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience recondary (thermal) stresses.
6.7 Pvametric Simulations A total of 21 dynamic simulations,19 Single Rack (SR) and 2 Whole Pool Multi-Rack (WPMR),
I have been performed to investigate the structural integrity of the fuel storage racks. These simulations cover a range of fuel loadings, coefficients of friction, and seismic loadings.
Single rack analyses were performed for the heaviest new rack (Rack P) and the new rack with ,
the highest aspect ratio (Rack L). Each rack was analyzed with three different fuel loadings, i.e.,
i
" fully loaded", " half loaded", and "nearly empty". The simulations also considered friction coefficients of 0.2, 0.5, and 0.8. The stress results from the SSE runs were conservatively compared with the OBE stress allowables, which are 50% less than the allowables for SSE. This approach eliminated the need to perform separate OBE simulations. l 1
One additional single rack run (Run No. 21) was carried out to satisfy the kinematic criteria in Subsection 6.6.1. This simulation considered an isolated fuel rack (i.e., no dynamic fluid coupling to the pool walls or adjacent racks was present). For this hypothetical case, only the displacement results were processed.
The rack numbering scheme for the Whole Pool Multi-Rack simulations is shown in Figure 6-1 together with the OCNGS letter designations. The model includes every rack in the Spent Fuel Pool, which numbers fourteen (4 new racks and 10 existing racks). The four new racks, which are identified by letter as racks L, M, N, and P, are assigned the numbers 14,13,10, and 9, respectively. This model was analyzed under SSE and OBE load conditions. For WPMR analysis, every rack was considered to be fully loaded with fuel. The coefficients of friction were generated based on a Gaussian distribution with a mean of 0.5 (with upper and lower limits of 0.8 and 0.2).
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-19 Report HI-981983
I The following table presents a complete listing of the simulations discussed herein.
l Run Seismic No. Model Type Fuel Loading Coefficient of Friction Event 1 WPMR SSE All racks are fully Gaussian distribution with a loaded. mean of 0.5 (upper aed lower j limits of 0.2 and 0.8) ;
2 WPMR OBE All racks are fully Gauuian distribution with a loaded. mean of 0.5 (upper and lower
)
limits of 0.2 and 0.8) )
3 SR (Rack P) SSE Fu!Iy loaded 0.2 4 SR (Rack P) SSE Fully loaded 0.8 5 SR (Rack P) SSE Fully loaded 0.5 6 SR (Rack P) SSE Halfloaded 0.2 7 SR (Rack P) SSE Halfloaded 0.8 8 SR (Rack P) SSE Halfloaded 0.5 9 SR (Rack P) SSE Nearly empty 0.2 10 SR (Rack P) SSE Nearly empty 0.8 11 SR (Rack P) SSE Nearly empty 0.5 12 SR (Rack L) SSE Fully loaded 0.2 13 SR (Rack L) SSE Fully loaded 0.8 14 SR (Rack L) SSE Fully loaded 0.5 15 SR (Rack L) SSE Halfloaded 0.2 16' SR(Rack L) SSE Halfloaded 0.8 17 SR(Rack L) SSE Halfloaded 0.5 18 SR(Rack L) SSE Nearly empty 0.2 19 SR (Rack L) SSE Nearly empty 0.8
, 20 SR (Rack L) SSE ,
Nearly empty 0.5 21 SR (Rack P) 3SE Fully loaded 0.8 Notes: SR = Single Rack; WPMR = Whole Pool Multi-Rack l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-20 Report HI-981983
I
/
s i
Run No. 21 was performed to evaluate the potential for overtuming of a single Holtec rack when isolated [in the pool without any fluid coupling to adjacent racks or walls. This simulation was performed to account for the unlikely possibility of a seismic event occurring during the installation process. In Run No. 21 the rack was fully loaded with fuel because increased fuel L loadings generally lead to higher displacements.
l l
l 6.8 Time History Results l
l 4
l l The results from the MR216 computer runs may be seen in the raw data output files. The 1 l
MR216 output files archive all of the loads and displacements, at key locations within each of the l 1
rack modules, at every time step throughout the entire time history duration. However, due to the !
l large quantity of output data, a post processor is used to scan the results for the maximum values and to calculate the stress factors, which are defined in Subsection 6.6.3.
Further reduction in this bulk of information is provided in this section by extracting the worst case results for the new racks; namely displacements, support pedestal forces, impact loads, and stress factors. This section also summarizes other analyses performed to develop and evaluate -
structural member stresses which are not determined by the post processor.
6.8.1 Rack Displacements The largest displacement is 0.469 inches which occurs in Run No. 21 for rack P in the X direction. The corresponding factor of safety against overturning is 105, which is much greater than the minimum required value of 1.5 for OBE conditions (see Subsection 6.6.1).
I SHADED TEXT CONTAINS PROPRIETARY INFORM ATION Holtec Intemational 6-21 Report HI-981983
f 6.8.2 Pedestal Vertical Forces l Pedestal number 1 for each rack is located in the +X, -Y corner of the rr.ck. Numbering increases t
l counterclockwise around the periphery of the rack. The highest pedestal load is 60,600 lb, and it occurs in Run No. I for pedestal 3 of rack M. Figure 6-13 provides a plot of the vertical pedestal l force, which is transmitted to the bearing pad, over the entire duration cf the event. l l
L 6.8.3 Pedestal Friction Forces i The maximum (x or y direction) shear load is 16,000 lb, and it occurs in Run No. I for rack P and is obtained by inspection of the complete tabular data.
6.8.4 Rack Imoact Loads A freestanding rack, by definition, is a structure subject to potential impacts during a seismic ;
event. Impacts arise from rattling of the fuel assemblies in the storage rack locations and, in l some instances, from localized impacts between the racks, or between a peripheral rack and the I pool wall. The following subsections discuss the bounding values of these impact loads.
6.8.4.1 Rack-to-Rack Imoacts I The single rack simulations do not predict any rack impacts. The WPMR results, however, do show impacts between adjacent racks at the baseplate level. There are no impacts at higher rack elevations. The maximum instantaneous impact loads are listed below.
Run Seismic Impact Load No. Model Type Locadon Event (Ib) 1 WPMR SSE 5,183 Between racks P and M at the baseplate level 2 WPMR OBE 4,604 Between racks P and M at j the baseplate level Notes: WPMR = Whole Pool Multi-Rack SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6 22 Report HI-981983 a
I
\
6.8.4.2 Rack-to-Wall Impacts The fuel racks do not impact the pool walls under any conditions.
6.8.4.3 Fuel to-Cell Wall Impact Loads
! A review of the results from each simulation allows determination of the maximum instantaneous impact load between fuel assembly and fuel cell wall at any modeled impact site.
The maximum fuel-to-cell wall impact is 226 lb, and it occurs in Run Nos. 7 and 8 for rack P.
The permissible lateral load on an irradiated spent fuel assembly has been studied by the Lawrence Livermore National Laboratory (LLNL). The LLNL report [6.8.1] states that "...for the most vulnerable fuel assembly, axial buckling varies from 82g's at inidal storage to 95g's after 20 years' storage. In a side drop, no yielding is expected below 63g's at initial storage to 74g's after 20 years' [ dry) storage." The most significant load on the fuel assembly arises from rattling during the seismic event. For the five lumped mass model, the limiting lateral load is equal to
- F = (w x a)/4 l
l where:
l l
w= weight of one fuel assembly (upper bound value = 690 lbf) a = permissible lateral acceleration in g's (a = 63)
Therefore, F. = 10,868 lbf. The maximum fuel-to-cell wall impact force is only 226 lbf, and the nominal factor of safety against fuel assembly yielding is 48.
I 1
i SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 6-23 Report HI-981983
l 6.9 Rack Structural Evaluation 6.9.1 Rack Dimensionless Stress Factors for Level B and D Loadings I
1 l The vertical and shear forces at the baseplate / pedestal interface are available as a function of 1 time. The maximum values for the stress factors defined in Section 6.6.3 can be determined for 1 every pedestal in the array of racks by scanning this data to select the limiting loads and performing calculations to determine member stresses. These two tasks are performed by a post-processor. With this information available, the structural integrity of the pedestal can be assessed and reported. Tlu vt section maximum (in time) bending moments and shear forces can also be i l determined at the baseplate / rack cellular structure interface for each spent fuel rack in the pool.
l This allows the evaluation of the maximum stress in the limiting rack cell (box).
l 6.9.1.1 Rack Pedestal Stress Factors The maximum pedestal stress factor is 0.362, which occurs in Run No. 2 for pedestal 3 of rack P.
This value is far less than the allowable !imit of 1.0, and therefore demonstrates compliance with ASME NF Code limits.
i 6.9.1.2 Rack Cell Stress Factors The stresses developed during SSE and OBE conditions remain below the allowable limit, and the rack modules are capable to withstand the loadings.
I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-24 Report HI-981983 i
r I
i l i
! 6.9.2 Pedertal Thread Shear Stress .
l \
l The largest thread shear stress computed by the post-processor is 3,631 psi under SSE conditions.
The yield stress and ultimate strength of the female por: ion of the pedestals are 21,300 psi and l 66,200 psi, respectively (Table 6-2). The male portion has much greater strength and is, therefore, not a controlling factor in the design. The allowable shear stress for Level B conditions is 0.4 times the yield stress, which gives 8,520 psi. The allowable shear stress for l Level D conditions is the lesser of 0.72 Sy(= 15,336 psi) and 0.42 Su (= 27,804 psi). Therefore, the former criterion controls. The computed stress of 3,631 psi is below the allowable shear stress for Level B. The thread shear stresses reported above are acceptable.
6.9.3 Local Stresses Due to Imoacts i
i Impact loads at the pedestal base (discussed in Subsection 6.8.2) produce stresses in the pedestal for which explicit stress limits are prescribed in the Code. The post-processor does not develop stress factors for the local regions of the racks that experience fuel to cell wall and rack to rack impact loads. Nese impact loads produce stresses that attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses.
So long as buckling does not occur, there are no limits on primary plus secondary stresses in the Code for Class 3 NF structures. Nevertheless, evaluations were made to ensure that the local impacts do not lead to plastic deformations in th: storage cells, which could affect the suberiticality of the stored fuel array.
- a. Impacts Between Fuel Assembly and Cell Wall Local cell wall integrity is conservatively estimated based on the peak fuel to cell wall impact load. Plastic analysis is used to obtain the limiting impact load, which leads to gross permanent deformation. Table 6-5 indicates that the limiting impact load of 260.5 Ib, which includes a safety factor of 2, is greater than the highest calculated impact load l of 226 lb (see Subsection 6.8.4.3) obtained from any of the rack analyses. Therefore, fuel
! ' impacts do not represent a significant concern with respect to fuel rack cell deformation.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-25 Report HI-981983
- b. Impacts Between Adiacent Racks i
i' As can be seen from Subsection 6.8.4.1, the bottom (baseplate) of the storage racks may l
impact each other during seismic events. Since the impact load is transmitted, on edge, between the 1-inch thick steel baseplates, the structural damage is expected to be '
negligible. If it is conservatively assumed that the impact load is spread over a 6-inch length of the baseplate, then the compressive stress that results from the highest impact
- load of 5,183 lb is only 864 psi, which is much less than the yield stress of the material.
Therefore, the predicted rack-to-rack impacts do not affect the configuration of the stored fuel.
y 6.9.4 Assessment of Rack Fatigue Marain Deeply submerged high density spent fuel storage racks arrayed in close proximity to each other i
in a free-standing configuration behave primarily as a nonlinear cantilevered structure when subjected to 3-D seismic excitations. In addition to the riulsations in the vertical load at each pedestal, lateral friction forces at the pedestal / liner interface, which prevent or mitigate lateral sliding of the rack, also exert a time varying moment in the baseplate region of the rack. The friction induced lateral forces act simultaneously in the x and y directions. The vectorial sum of the x and y direction forces can not exceed V, where is the limiting interface coefficient of friction and V is the concomitant vertical thrust on the liner (at the given time instant). As the vertical thrust at a pedestal location changes, so does the maximum friction force, F, that the interface can exert. In other words, the lateral force at the pedestal / liner interface, F, is given by l F s N(t) where N (vertical thrust) is a function of t (time). F does not always equal N; rather, N is the l maximum value it can attain at any time; the actual value, of course, is determined by the l
dynamic equilibrium of the rack structure. In summary, the horizontal friction force at the pedestal / liner intedace is a function of time; its magnitude and direction of action vary during the l
earthquake event.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-26 Repo;t HI-981983
The time varying horizontal and vertical forces on the extremities of the support pedestals produce stresses at the root of the pedestals in the manner of an end loaded cantilever. The stress field in the cellular region of the rack is quite complex, with its maximum values located in the region closest to the pedestal. The maximum magnitude of the stresses depends on the severity of the pedestal end loads and on the geometry of the pedestal / rack baseplate region.
Altemating stresses in metals produce metal fatigue if the amplitude of the stress cycles is sufficiently large. In high density racks designed for sites with moderate to high seismic action, the stress intensity amplitudes frequently reach va: lues above the material endurance limit, leading to expenditure of the fatigue " usage" reserve of the material.
Because.of the locations of maximum stress (viz., the pedestal / rack baseplate junction) and the close placement of racks, a post-earthquake inspection of the high stressed regions in the racks is not feasible. Therefore, the racks must be engineered to withstand multiple earthquakes without reliance of nondestructive inspections for post-earthquake integrity assessment. The fatigue life evaluation of racks is an integral aspect of a sound design. l The time history method of analysis, deployed in this report, provides the means to obtain a !
complete cycle history of the stress intensities in the highly stressed regions of the rack. Having j determined the amplitude of the stress intensity cycles and their number, the cumulative damage factor, U, can be determined using the classical Miner's rule, U=I E Ni where ni is the number of stress intensity cycles of amplitude oi, and Ni is the permissible number of cycles corresponding to ai from the ASME fatigue curve for the material of construction. U must be less than or equal to 1.0.
To evaluate the cumulative damage factor, a finite element model of a portion of the spent fuel rack in the vicinity of a support pedestal i- snstructed in sufficient detail to provide an accurate SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-27 Report HI-981983
I l
assessment of stress intensities. Figure 6-14 shows the essentials of the finite element model.
The firite element solutions for unit pedestal loads in three orthogonal directions tre combined to
)
establish the maximum value of stress intensity. Using the archived results of the spent fuel rack dynamic analyses (i.e., pedestal load histories versus ti.me), enables a time history of stress intensity to be developed at the most limiting location. This establishes a set of alternating stress intensities ranges versus number of cycles for several seismic events. Following ASME Code guidelines, it is found that U = 1.47 x 10 2 for the combined load of I SSE and 20 OBE events. !
l This cumulative damage factor is below the ASME Code limit of 1.0, and therefore fatigue failure is not expected.
1 6.9.5 Weld Stresses I Weld locations subjected to significant seismic loading are at the bottom of the rack at the
' baseplate to cell connection, at the top of the pedestal support at the baseplate connection, and at cell to cell connections. Bounding values of resultant loads are used to qualify the connections.
- a. Baseolate to Cell Welds Reference [6.6.1] (ASME Code Section III, Subsection NF) permits, for Level A and B conditions, an allowable weld stress of 0.3 Su (= 19,860 psi). As stated in Subsection 6.6.2, the allowable weld stress for Level D conditions is 0.4 Su, which equals 26,480 psi.
Weld dimensionless stress factors are determined by applying a simple. conversion (ratio) factor to the corresponding cell stress factor. The factor is equal to the ratio of the cross sectional area of the cell metal to the stress area of the cell-to-baseplate weld. The ratio equals
, 0.075in x (5.93in + 0.075/n)
= 2.04 0.0625in x 0.7071 x Sin The highest predicted weld stress for SSE is calculated from the maximum cell stress.
The final result is less than the OBE allowable weld stress, which is 19,860. Therefore, all weld stresses between the baseplate and the cell walls are acceptable.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-28 Report HI-981983
1 l
l
- b. Baseolate-to-Pedestal Welds The welds between the baseplate and support pedestal are evaluated by development of a finite element model of the bearing pad / base plate interface and appropriate application of the maximum pedestal loads. The maximum weld stress was determined to be 7,587 psi, which is much less than the OBE allowable weld stress of 19,860 psi. The results are also shown in Table 6-5.
- c. Cell-to-Cell Welds 1
Cell-to-cell connections are formed by a series of welds along the cell height. Stresses in the celi to cell welds develop due to fuel assembly impacts with the cell wall and the bending motion of the rack (i.e., shear flow).
- The weld stresses due to fuel impacts are conservatively calculated by assuming that fuel assemblies in adjacent cells move out of phase so that the impact loads in two adjacent I cells are in opposite directions; this tends to separate the two cells from each other at the weld. Table _6-5 gives results for the maximum allowable impact load that can be transferred by the cell-to-cell welds. An upper bound on the load required to be transferred is also given in Table 6-5, and it is much lower than the allowable load. This upper bound value is conservatively obtained by applying the bounding rack to fuel impact load from any simulation in two orthogonal directions simultaneously and multiplying the result by 2 to account far the simultaneous impact of two assemblies. An equilibrium analysis at the connection then yields the upper bound load to be transferred.
It is seen from the results in Table 6-5 that the calculated load is well below the allowable.
The maximum stress in the cell-to-cell welds due to shear flow is 16,861 under SSE conditions. This result is less than the OBE allowable weld stress of 19,860 psi. The maximum stress develops in the lowermost cell-to-cell weld, which is adjacent'to the pedestal.
6.9.6 Bearina Pad Analysis
. To protect the pool slab from high localized dynamic loads, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact these bearing pads during a seismic event, and the pedestal load is transferred to the slab. Bearing pad dimensions are chosen to ensure that the average pressure on the slab surface, due to a static load plus a dynamic impact load, does not exceed the American Concrete Institute, ACI-349 (6.9.1] limit on bearing pressures. Section 10.15 of[6.9.2] gives the design bearing strength as SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-29 Report HI-981983
l l
J f3= $(0.85 fe')c (
where $ = 0.7 and fe' is the specified concrete strength for the spent fuel pool. c = 1, except ;
.when the supporting surface is wider on all sides than the loaded area. In that case, c =
(A 2/Ai)", but not more than 2. A isi the actual loaded area, and A is 2 an area greater than A i, i which is defined in [6.9.2). Using a value of e > 1 includes credit for the confining effect of the surrounding concrete. It is noted that these criteria are in conformance with the ultimate strength primary design methodalogy of the American Concrete Institute in use since 1971. For the spent fuel pool at Oyster Creek, the concrete compressive strength is fe' = 3,000 psi. The allowable .
bearing pressure is conservatively computed by taking c = 1 to account for lack of total concrete confinement in the leak chase region and a stress reduction factor of $ = 0.7, Thus, the maximum allowable ediicrete bearing pressure is 1,785 psi.
1 l
fhe maximum vertical pedestal load is 60,600 lb. The bearing pad is 2" thick, austenitic I
stainless steel plate stock. The average pressure at the pad to liner interface is computed and compared against the above-mentioned limit. Calculations show that the average pressure at the slab interface reaches a maximum of 1,534 psi, which is below the allowable value of 1,785 psi.
This yields a safety factor of 1.16. Therefore, the bearing pad design devised for the Oyster Creek spent fuel pool is deemed appropriate for the prescribed loadings.
6.9.7 Level A Evaluation The Level A condition is not a goveming condition for spent fuel racks since the general level of loading is far less than the Level B loading. To illustrate this, the heaviest of the new spent fuel racks is considered under the dead load. It is shown below that the maximum pedestal load is low and that further stress evaluations are unnecessary.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-30 Report HI-981983
(
l l
LEVEL A MAXIMUM PEDESTAL LOAD l
Dry Weight of Largest New Rack = 9,700 lb l Dry Weight of 107 BWR Fuel Assemblies = 73,830 lb l Total Dry Weight = 83,530 lb l 1 Total Buoyant Weight (0.87 x Total Dry Weight) = 72,671 lb I Load per Pedestal = 18,168 lb' The stress allowables for the normal condition is the same as for the upset condition. The maximum pedestal load during upset conditions is 54,600 (see Subsection 6.8.2). Since this load (and the corresponding stresses in the rack) is much greater than the 18,168 lb load calculated above, the upset (OBE) condition controls over the normal condition. Therefore, no further I evaluation is necessary for Level A.
l 6.10 Hydrodynamic Loads on Pool Walls
)
l The maximum hydrodynamic pressures that develop between the fuel racks and the spent fuel pool walls are calculated for each Whole Pool Multi-Rack simulation. The maximum hydrodynamic pressures for the SSE and OBE events are 3.25 psi and 2.5 psi, respectively.
These hydrodynamic pressures are considered in the evaluation of the spent fuel pool structure.
l h
1 I
l l
This load is based on four pedestals per rack. The rack that is considered actually has five pedestals.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-31 Report HI-981983 t
L 1
6.11 Conclusions o
l e Time history simulations, including all nonlinear impact and interface friction effects, have been employed to evaluate the structural margins in the Holtec spent fuel racks.
i e The totality of simulations provide an extensive set of results for loads, stresses, and displacements, which taken together demonstrate that the spent fuel racks meet the l governing Code requirements.
l e Evaluation of structural margins havc been performed for the array of racks in the spent fuel pool under various conditions (e.g., fuel loading, coefficient of friction, etc.). The requirements of the specification and the governing Code documents are met for Level A, Level B, and Level D conditions.
Based on all of the results presented in Section 6, the spent fuel racks are qualified for the intended service.
6.12 References for Section 6
[6.1.1] USNRC NUREG-0800, Standard Review Plan, June 1987.
l
[6.1.2] (USNRC Office of Technology)"OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978, l and January 18,1979 amendment thereto. 1
[6.2.1] Soler, A.I. and Singh, K.P., " Seismic Responses of Free Standing Fuel l Rack Constructions to 3-D Motions," Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984).
[6.2.2] Soler, A.I. and Singh, K.P., "Some Results from Simultaneous Seismic Simulations of All Racks in a Fuel Pool," INNM Spent Fuel Management Seminar X, January,1993.
[6.2.3] Singh, K.P. and Soler, A.I., " Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the. Chin Shan Experience," Nuclear '
Engineering International, UK (March 1991).
[6.2.4] Holtec Proprietary Report HI-961465 - WPMR Analysis User Manual for Pre- & Post-Processors & Solver, August 1997.
[6.4.1] NUREG-0800, Standard Review Plan 3.7.1, " Seismic Design Parameters,"
Rev. 2, August 1989.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-32 Report HI-981983
[6.4.2] Holtec Proprietary Report HI-89364 - Verification and User's Manual for Computer Code GENEQ, January,1990.
l [6.4.3] EQE International, Calculation 100028-C-01, Revision 0, April 9,1997.
I
[6.4.4] NUREG-0800, Standard Review Plan 3.7.2, " Seismic System Analysis,"
l Rev. 2, November 1989.
[6.5.1] Rabinowicz, E., " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company,1976.
[6.5.2] Singh, K.P. and Soler, A.I.," Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium: The Case of Fuel Racks," 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.
[6.5.3] Fritz, R.J.,"The Effects of Liquids on the Dynamic Motions ofImmersed Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172.
[6.5.4] Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering,"
McGraw Hill,1976.
[6.5.5] Paul, B., " Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment (Proprietary)," NUSCO/Holtec Report HI-88243.
[6.6.1] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF,1995 Edition.
[6.6.2) ASME Boiler & Pressure Vessel Code,Section II, Part D,1995 Edition.
[6.6.3] USNRC Standard Review Plan, .NUREG-0800 (Section 3.8.4, Rev. 2, 1989).
[6.6.4] ASME Boiler & Pressure Vessel Code,Section III, Appendices,1995 Edition.
[6.6.5] USNRC Regulatory Guide 1.124, " Service Limits and Loading
- Combinations for Class 1 Linear-Type Component Supports," Rev.1, l January 1978.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-33 Ileport HI-981983 J
I
[6.8.1] Chun, R., Witte, M. and Schwartz, M., " Dynamic Impact Effects on Spent ,
Fuel Assemblies," UCID-21246, Lawrence Livermore National Laboratory, October 1987 l [6.9.1] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan,1985.
[6.9.2] ACI 318-95, Building Code Requirements for Structural Concrete, j American Concrete Institute, Detroit, Michigan,1995.
i l
l l
l i
I l
t l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6 34 Report HI-981983
r l
Table 6-1 i
. l PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK l
l l PLANT DOCKET NUMBER (s) YEAR Enrico Fe mi Unit 2 USNRC 50-341 1980 l Quad Cities I & 2 USNRC 50-254,50-265 1981 Rancho Seco USNRC 50-312 1982 I
Grand Gulf Unit 1 USNRC 50-416 1984 Oyster Creek USNRC 50-219 1984 Pilgrim USNRC 50 293 1985 V.C. Summer USNRC 50-395 1984 Diablo Canyon Units 1 & 2 USNRC 50-275,50-323 1986 Byron Units 1 & 2 USNRC 50-454,50-455 1987 l l Braidwood Units 1 & 2 USNRC 50-456,50-457 1987 Vogtle Unit 2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 Millstone Point Unit 1 USNRC 50-245 1989 Chinshan Taiwan Power 1988 D.C. Cook Units 1 & 2 USNRC 50-315,50-316 1992 Indian Point Unit 2 USNRC 50-247 1990 Three Mile Island Unit 1 USNRC 50-289 1991 l James A. FitzPatrick USNRC 50-333 1990 i Shearon Harris Unit 2 USNRC 50-401 1991 Hope Creek USNRC 50-354 1990 1
7 Kuosheng Units 1 & 2 Taiwan Power Company 1990 l Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units 1 & 2 Comision Federal de Electricidad 1991
_ Zion Station Units 1 & 2 USNRC 50 295,50-304 1992 SHADED TEXT CONTAINS PROPRIETARY INFORMATION ,
Holtec International 6-35 Report HI-981983 I
l l
l l
i l
Table 6-1 i l
l PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK l PLANT DOCKET NUMBER (s) YEAR Sequoyah USNRC 50-327,50-328 1992 LaSalle Unit i USNRC 50-373 1992 Duane Arnold Energy Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit 1 USNRC 50 220 1993 Beaver Valley Unit 1 USNRC 50-334 1992 Salem Units 1 & 2 USNRC 50-272,50-311 1993 Limerick USNRC 50-352,50-353 1994 Ulchin Unit 1 KINS 1995 Yonggwang Units I & 2 KINS 1996 Kori-4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 Angra Unit i Brazil 1996 Sizewell B United Kingdom 1996 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-36 Report HI-981983
1
)
Table 6-2 RACK MATERIAL PROPERTIES (200*F)
(ASME - Section II, Part D)
Young's Modulus Yield Strength Ultimate Strength Material E Sy Su (Psi) (psi) (psi) 6 SA240-304L 27.6 x 10 21,300 66,200 6
SA564-630 (male portion of 28.5 x 10 106,300 140,000 pedestal; age hardened at 1100*F) 1 I
1 l
l l
l i
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-37 Report Hi-981983 L .
Table 6-3 DEGREES OF FREEDOM i i
LOCATION (Node) DISPLACEMENT ROTATION Ux Uy U, 0x Oy 0, ;
1 Pi P2 P3 q4 qs 96 2 P7 P P9 gio gii gi2 Node 1 is assumed to be attached to the rack at the bottom most point.
Node 2 is assumed to be attached to the rack at the top most point.
Refer to Figure 6-8 for node identification. l 2' p3 pi4 3' pis pi6 +
4' pi7 pia 5' pi, p20 l 92: p22 where the relative displacement variables qi are defined as:
pi =
qi(t) + Ux(t) i = 1,7,13,15,17,19,21
=
qi(t) + Uy(t) i = 2,8,14,16,18,20,22 !
= q,(t) + U,(t) i = 3,9 l
=
q,(t) i = 4,5,6,10,11,12 l
pi denotes absolute displacement (or rotation) with respect to inertial space qi denotes relative displacement (or rotation) with respect to the floor slab
- denotes fuel mass nodes U(t) are the three known earthquake displacements l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-38 Report HI-981983
i Table 6-4 (bYNARACK) NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICT ELEMENTS I. Nonlinear Springs (Type 3 Gap Elements - 488 Total)
Number Node Loc. Description 1 Support S1 Z compression-only Jcment 2 Support S2 Z compression-only element 3 Support S3 Z compression-only element 4 Support S4 Z compression-only element 5 2,2* X rack / fuel assembly impact element between nodes 2 and 2*
6 2,2* X rack / fuel assembly impact element between nodes 2 and 2' 7 2,2* Y rack / fuel assembly impact element between nodes 2 and 2*
8 2,2* Y rack / fuel assembly impact element between nodes 2 and 2*
9-24 Rack / fuel assembly impact elements attached to the rattling masses at nodes l',
3*,4* and 5* (similar to element numbers 5 through 8)25-336 Spring elements I through 24 are repeated for rack numbers 2 through 14 337-488 Rack periphery Inter-rack impact elements at top elevation and at baseplate elevation l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION l Holtec International 6 39 Report HI-981983 l
F Table 6-4 l (DYNARA CK) NUMBERING SYSTEM FOR GAP EI.EMENTS AND FRICTION l
ELEMENTS l
l l
II. Linear Springs (Type 1 Elements - 84 Total)
Number Rack No. Description 1 I Rack beam bending element (x-z plane) 2 1 Rack shear deformation element (x-z plane) 3 1 Rack beam bending element (y-z plane) 4 1 Rack shear deformation element (y-z plane) 5 1 Rack beam axial deformation element 6 1 Rack beam torsional deformation element 7-84 2-14 Spring elements 1 through 6 are repeated for rack numbers 2 through 14 III. Piece-wise Linear Friction Springs (Type 2 Elements - 112 Total)
Number Rack No. Description 1 1 Pedestal 1, X direction 2 1 Pedestal 1, Y direction l 3 i Pedestal 2, X direction
)
4 1 Pedestal 2, Y direction 1
5 1 Pedestal 3, X direction i 6 1 Pedestal 3, Y direction 7 1 Pedestal 4, X direction l 8 1 Pedestal 4, Y direction 9-112 2-14 Spring elements 1 through 8 are repeated for rack numbers 2 through 14 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 6-40 Report HI-981983
F l
l l
Table 6-5 l
l l COMPARISON OF BOUNDING CALCULATED LOADS / STRESSES VS.
CODE ALLOWABLES l AT IMPACT LOCATIONS AND AT WELDS l SSE I
Item / Location Calculated Allowablet i
Fuel assembly / cell wallimpact,Ibf 226 260.5" Baseplate / pedestal weld, psi 7,587 19,860 Cell / cell weld,Ibf 639"' 3,195 l
l i
i i
i I
Note that Level A condition allowables are conservatively applied against SSE loads.
Based on the limit load for a cell wall.
"t Based on the maximum fuel assembly-to-wall impact of 226 lb simultaneously applied in two orthogonal directions.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-41 Report HI-981983
l l
\ -
/ =
=
- GH =
= OK _
I f I -
. - II . _
S I _
il f :. ll , l S -
Y
- ; = L
[ I t = A N
_. Og
,L
- r! t A
@j d_ l =
=
R M
@M _
P
@P _ =
=
W l l l I_ li!l u .l.. l l l , j R
O
_ _ F
@L _
_ ON _
. E
=
=
M E
c 5' ,. l ; L I I
C .
_ =
S O0 =
@F G N
@B I R
=
=
E B
M U
!,l:I. l il ,, : l N
- K C
= _
- =
_ A _
R
- 1
- @A =
=
@E 6
E R
=
=
U G
I
)l
- : . , l - F l l i
m m
i m
3 8
0 9 1
2 8 9-I I
I N~
t r
o p
e 8
1 R
l l'
Mi" l
l L E 6 1
ii9 _
i 4
1 _
lI l
l l y
ll l
l iO I l
q I
r o
t s ' 2 i 1 l
l.
H Il Il i
T t
n a
0 1
c e
s e
2-6 r m e e i r
l e T u c i g _
c i A i' F k
e 8 e
r l C
t r
e s
l l
Y l -
y G
6 i
l[
l l
n ll 4
I l l l _
ll 3
2 l
l i a n
o
/
t' 0 i
t a
n r
e 0 0 0 0 0 0 t n
5 3 , 3 5 7 I
~ - - -
c e
ij l t
o
_ I I
3 8
0 9 1
2 8 9-1 1
1 t
r o
p d e k
i 8
1 R
b _
i 6 _
1 _
"r 1 _
i .
A '
p l
4 1 .
l n
I y
i I
I 2 _
1 l
n e c e
i s
0 , 3-1 e 6
Y m i e t
n n
I t T u r
g e
v i E
E 1' F B 8 O 1 i
' I 6
f l
i -
i 4
1
'[
_ l I
2 l
a n
f i o
t a
A_
0 n r
e 0 0 t
0 0 n 5 3 1 5 I c
e j l t
o I
t I
3 8
0 9 1
2 8
. 9-I H
t r
o p
e 8 R E
1 l
Jl bl I 6 1
i l
i E'
l
.l l
ll li p 4 1
l il _
ll li
' l _
HE
' li l _
" l 2 .
' J l l
1 _
k c a1 t, I1 i i
n agN l _
1i l
l l
0 4- _
1 _
l l 6 .
c l e
r
' u g
I U i
u: F Ii E
c Ll l c
l
'l
's Il _
B
' I c
' I
" Il I
y _
hgV I l
lN i
l f _
l l _
l 4 _
ldl _
IH I lIl _
li J
__ I _
l
,E l .
l l l l l 2 -
l lI I
l
. a n -
i o
t _
a
.P i
O n r
I t
e n
c e
e l t
l o .
l _
e 8
=
l t 8.
e
- N ,
4 5
. N
===mia lllll;;;:""
mqm -
- g-
= :===l E
~
g -
3 M=_ == - _. . . 1 s
"=iEi =" = =- m 4 i g -
===e "=::=!i.-
g i
E ---
f I === C5iiii!!
E: I 4-
==
. =::-
- =_ n===--
2E ge; _
mummi r_ _
=giML _
m_==== ===i!in== rl ='""l:=!-
=. =- - R
-- ::g;;;g .Allii== =: .
g
~. -=== 7 =m ==:. m C.
~ 5;.
<=,
4 -
a a e a =
g w ::m 9 , ,
.)
g 5
8 i
=
4 I
I 8 "
h I
l l
l S
4 4~
w m
N w
==
H ~
--_g _
yl}
__===
, _ _- g
=mmmm-man N ~ .,
w-1-
{ -
m
_ ===i m
~
~
e
-==--m==== ,
amma W *
"""""""" sm _- -
- M i - "mm"mm"m "m"e"m"*.
t 1 i i a e a =
u.:mme=
a 9 4 j 3
3_i
a e
eo d
I =
e b \
u e
i 4 i
1 8 !
E ~
- t l
l -
l ,
1 l
r n o
o - -
o B !
.ee m i r i t.
- d. ,
4 i
i i
I 1
R
)
i 1ll 8 s e a a a
iO c
O 9 g g
~
8
=
0 t
.)
u q22 l RA TTLING n Pl9
/ FUEL MASS
( TYP. )
e i
TCP GF RACK GAP -
MENTS g p,g g
q, 017 _ _ _ _
q20
' ~"
'l , ./
RACK GE0 METRIC CENTER LINE
{f,/ pio k9 LINEAR BEAM SPRING (6 TOTAL ) g4+ 4,,
/
PI2 kII q6 PEDESTAL 3
/
%_ 5+ 7 / /
PEDESTAL 4 P3n&Lsy ' -
rI4
$13 4l
?
j P2 n qf _
PEDESTAL 1 '
Pit---- ?/ ' 15
/ /," PEDESTAL 2 q4 f PIS /
, s
=f FRICTION uV.
\
uf PEDESTAL /FOUNDA TION ELEMENT X GAP ELEMENT FIGURE 6-8 SCHEMATIC 0F THE DYNAMIC MODEL OF A SINGLE RACK MODULE USED IN DYNARACK REPORT # HI-981983l lHOLTECPROPRIETARY
,+ 4
v Y
R -
A -
S T E
LT S I -
LC A R EA P AP M M O YI R L P T BT G -
N A MN C -
EE E
_ L SM I T
. W L T H
_ LV EE FG A _
CL R -
gj F
O L
E -
V x E -
L
=
T A
/ M -
E
/ L E
_ f / T
/ C _
Y A -
/
P -
M
_ /
_ /
- ' I
/
=
P -
A -
G -
K -
C A
R -
O T
L g E X U F
/ Y B
9 6
E R
U l 3
= G 8 I 9
\ F 1 8
9 s I l
H
I FUEL ASSY./ CELL CAP /IWPACT ELEMENT, Ki
. Aom ;
H/4 H/2
)N TYPICAL FUEL i RACK C.C. - RATTLING MASS I W4 RACK PERIPHERY -
S b) M ,
CAP /IWPACT !
ELEWENTS, Kg I
H/4 I ,
b%l[)M' "
H/2 - --
V4 p
wA CAP / IMPACT FRICTION
- ELEMENT.KS INTERFACE ELEMENT,Kf N /7 77 FIGURE 6-10 TWO DIMENSIONAL VIEW OF THE SPRING-MASS SIMULATION REPORT # HI-981983l lBOLTECPROPRIETARY
(;
, 98 il E
n H/2 fAq10 YO T
Ksr "*
g#
H/2 C
RACK DEGREES-OF-FREEDOM FOR Y-2 PLANE BENDING WITH SHEAR AND BENDING SPRING it I
h H/2 f%
Y II Ksx N "'
H/2 C
95 FIGURE 6-11 RACK DEGREES-OF-FREEDOM FOR X-2 PLANE BENDING WITH SHEAR AND BENDING SPI'!NG HI-981983l IHOLTEc PRopgig7ggy
f a
b
'l
/ "^jkET /
/ /I /
0 TYPICAL TOP IMPAET ELEMENT RAEK STRUETURE IMPAET ELEMENT
/ / /
RAEK PLATE 1 P FIGURE 6-12 RACK PERIPHERY GAP / IMPACT ELEMENTS l
)
REPORT # HI-981983 HOLTEC PROPRIETARY l
_ 3 8
l 0 9 2 1 8
i 9-I H
}
g
_ 8 1
I d
\
I 6
l 1
i l
l I t I
. 4 1
i n
i o
t 2 t
a e 3 ' 1 Sg mla i
l nTt s i . se t
- a r vd e e -
l r e e
n ceP l s
e r , 0 e 3
G FoM I 1 1 r
m 6 al e a kc '
Ti e r
s a t
l c eR u g
- ud .
l- lll i N e1 F .
P . l k
e l o 8 e
r ci aN ) .
Cn n u r e eV R t
s l .
y O I 6
l
- [al ,
I 4
o I l
l '
2 _
h Y'
0 0
^ 0 0 0 0 0 0 0 0 0
0 7
0 0
6 0
0 5
0 0
3
~
1 -
E. 35$a2e i
i E
5 b -
1 ,
j ' _
~~
N a
m u
2 h
LE
[ , .
I l
p l- 7.0 FUEL HANDLING AND CONSTRUCTION ACCIDENTS l .
.7.1 -Introduction The USNRC OT position paper [7.1.1) specifies that the design of the rack must ensure the l
functional integrity of the spent fuel racks under all credible fuel drop events in the spent fuel pool.
This section contains synopses of the analyses carried out to demonstrate the regulatory compliance .
of the proposed racks under postulated fuel assembly drop scenarios germane to Oyster Creek.
In addition to the postulated fuel assembly drops, an accident event daing the rack installation is
. evaluated in which the heaviest rack falls from the surface of water. In this case, the ability of the pool structure to avert primary structural damage (leading to rapid loss of water) must be demonstrated.
7.2 Fuel Handline Accidents 7.2.1 Description For the fuel handling accidents discussed herein, the concem is with the damage to the storage racks, the fuel assemblies, and the pool structure. The positions of the fuel assemblies, rack cells, and neutron absorber material must remain consistent with the configurations used in the criticality evaluations. The results of the criticality evaluations remain valid if the general configuration is maintained.
Radiological concems due to fuel damage are not an issue for the proposed change, since the previously evaluated design basis fuel handling accident considers the worst case condition, which remains unchanged. The worst case condition is a fuel assembly falling to the floor and striking the liner. Fuel damage subsequent to a drop is primarily influenced by the weight and design of the fuel assembly, the drop 11eight (which determines the kinetic energy upon impact), and the orientation of the falling assembly. Since none of these parameters are changed under the proposed modification, the number of fuel rods damaged during a fuel assembly drop remains consistent with the previously analyzed fuel handling design basis accident.
l SHADED TEXT CONTAINS PROPRIETARY INFORM ATION l Holtec Imernational 7-1 Report 111981983
Two types of drops are evaluated to consider the impact zone at the top of the racks and at the baseplate. The first class considers a falling fuel assembly striking the top of the storage racks. The kinetic energy of the falling assembly is absorbed by damage to the racks. In this so-called " shallow ,
drop" event, a fuel assembly, along with the portion of handling tool that is severable in the case of a single element failure, is assumed to drop vertically and hit the top of the rack. Inasmuch as the new racks are of honeycomb construction, the deformation produced by the impact is expected to .
be confined to the region of collision. However, the " depth" of damage to the affected cell walls must be limited to the portion of the cell above the top of the " active fuel region", which is essentially the elevation of the top of the Boral neutron absorber. To meet this criterion, the plastic deformation of the rack cell wall should not extend more than 18 inches (downwards) from the top 1 of the rack. This will en;ure that the configurations considered in the criticality evaluations are not compromised.
1 The postulated drop event occurs during the manipulation of an 800 lb load comprised of a fuel l i
assembly and its handling tool. The origin of the drop trajectory is 36 inches above the rack, which is chosen to. bound the highest elevation that the load can be lifted by the spent fuel handling machine. The impactor travels through the stratum of water before impacting the upper portion of the rack. For conservatism, the analysis considers a 1500 lb bounding drop load.
It is readily apparent from the description of the rack modules in Section 3 that the impact resistance
- of a rack at its periphery is less than its interior. Accordingly, the potential shallow drop scenario is postulated to occur at a rack periphery cell in the manner shown in Figure 7-1. The impact zone is chosen to minimize the cross sectional area that experiences the deformation.
In order to maximize the penetration into the top of the rack by the falling assembly, the rack is considered empty. Exclusion from the model of fuel stored in cells adjacent to the impact site eliminates the possibility of secondary fuel impacts, thus maximizing rack damage (i.e., depth of penetration). The impact scenario, therefore, emphasizes the magnitude and the extent of the plast:c deformation of the'0.075 inch peripheral wall of the comer cell.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 7-2 Report HI-98!983 3
l
The second class of event postulates that the impactor falls through an empty storage cell impacting the rack baseplate. This so-called " deep drop" scenario threatens the structural integrity of the baseplate and the pool liner. If the baseplate is pierced, then the fuel assembly might damage thq.
pool liner and/or create an abnormal condition in which the enriched zone of the fuel assembly is outside the " poisoned" space of the fuel rack. To preclude damage to the pool liner and to avoid the potential for an abnormal fuel storage configuration in the aftermath of a deep drop event, it is .
required that the baseplate remains intact and that the maximum deflection of the fuel assembly i
support surface is less than the distance from the bottom of the rack baseplate to the liner. l The deep drop event can be divided into two cases, namely, a drop through a cell located above a support leg and a drop through an interior cell away from the support pedestal (see Figure 7-2). In the former case, the baseplate is buttressed by the suppon pedestal and presents a hardened impact surface, resulting 'in a high impact load. The principal design objective is to ensure that the support pedestal does not pierce the spent fuel pool liner. At cell locations away from the pedestal, the baseplate is not as stiff. In this case, baseplate severing and large deflection of the baseplate (such that the liner would be impacted) are unacceptable results.
7.2.2 Mathematical Model In the first step of the solution process, the velocity of the dropped object (impactor) is computed !
for the condition of undenvater free fall. Table 7-1 contains the velocity results for the three drop events.
1 1
In the second step,'an elasto-plastic finite element model of the impacted region is prepared on Holtec's computer code PLASTIPACT (Los Alamos Laboratory's DYNA 3D implemented on Holtec's QA system). PLASTIPACT simulates the transient collision event with full consideration ,
of plastic, large deformation, wave propagation, and elastic / plastic buckling modes. For conservatism, the impactor in all cases is conservatively assumed to be rigid. The material properties and the structural definitions of the impactor and the target are summarized in Tables 7-2 and 7-3 for each impact event.
- SHADED TEXT CONTAINS PROPRIETARY INFORMATION l Holtec International 7-3 Report III-981983
7.2.3 - Results 7.2.3.1. Shallow Drop Events- ;
i
, l Figure 7-3 provides an isometric view of the finite element model utilized in the shallow drop impact analysis.
i Dynamic analyses show that the top of the impacted region undergoes localized deformation. Figure 7-4 shows an isometric view of the post-impact geometry of the rack for the shallow drop scenario, as well as a contour plot of the Von-Mises stresses. The maximum depth of plastic defonnation is limited to 7.43 inches, which is below the design limit of I8 inches. Figure 7-5 shows the plan view of the post-impact geometry. Approximately 40% of the cell opening in the impacted cell is blocked. The SFP locnl temperature evaluation in Subsection 5.6.2 conservatively assumes 50%
blockage at every celllocation.
I 7.2.3.2 Deep Drop Events A plan view of the finite element model for the " deep drop" condition through an interior cell is shown in Figure 7-6. The collision between the 1500 lb impactor and the rack baseplate occurs at a velocity of 357 in/sec, and it results in an area oflocal deformation, which extends over a 15 inch square area. The baseplate does not fracture during the impact, but the welds connecting the adjacent cells to the baseplate are severed. The maximum deflection of the baseplate is 2.26 inches, which is less than the distance of 8.1875 inches from the baseplate to the liner. The maximum calculated Von Mises stress and plastic strain in the baseplate are 44.3 ksi and 0.106, respectively. Figure 7-7 shows the deformed shape of the baseplate. The effect of the baseplate deformation, due to a dropped fuel assembly, on the criticality safety analysis is considered in Subsection 4.7.5.
The " deep drop" scenario, in which the impact. occurs above the support pedestal, produces '
negligible baseplate deformation. The 1/4 inch pool liner is not pierced during the collision, since the maximum Von Mises stress of 27.3 ksi is less than the failure stress of 71 ksi for the liner material, which is given in Table 7-2.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 7-4 Report III 981983
Figure 7-8 shows an elevation view of the finite element model for the case where the impact is above a support pedestal. A high stress (48 ksi) is observed in the pedestal cylinder at the contact surface with the bearing pad. However, this value is below the failure stress of 140 ksi (Table 7-21 for this material. The bearing pad registers a maximum Von Mises stress of 22 ksi located in the contact area. This stress is well below yield for this material and is, therefore, acceptable. The concrete stratum directly beneath the pedestal sustains a very localized (peak normal) compressive .
1 stress with a maximum amplitude of 278 psi, as shown in Figure 7-9. This results in localized damage to the concrete (i.e., experiences only localized crushing). The rest of the modeled area is in tension, but the stress is only 241 psi, which can be supported by the concrete without cracking.
7.3 Rack Dron The drop of a rack during the rack installation is also postulated. This evaluation considers an empty fuel rack, which is dropped to the bottom of the pool from a height of 40 feet. The damage to the liner and underlying concrete is determined by neglecting any bearing pads at the impact site and considering that the pedestal directly strikes the unprotected liner. The results indicate that the liner is pierced and that the pedestals indent the concrete by approximately 2.7 inches. Local cracking is also observed in the concrete material.
7.4 Conclusions The postulated fuel assembly drop accidents are analyzed and found to produce localized damage well within the design limits for the racks. The post-impact configuration of the fuel assemblies and neutron absorber material (Boral) is consistent with the configuration analyzed in the criticality evaluations discussed in Section 4.0. The calculated baseplate deformation and the corresponding fuel displacement are considered in the criticality evaluations. There are no further criticality concems due to these accidents. The damage to the top of the racks reduces the cross sectional area available for coolant flow. The reduction of area, however, is less than what is considered in the thermal-hydraulic evaluations. The rack drop analysis shows that the pedestals pierce the pool liner.
The loss of spent fuel coolant due to potential linte leakage, however, was addressed in the Oyster Creek design basis V installing a leakage detection system beneath the liner and by providing the SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 7-5 Report 111981983
capability to make up the loss ofinventory to the pool This is consistent with SRP 9.1.2 and SRP 9.1.3. ihe concrete penetration is less than the cover depth of the steel reinforcement. Therefore, the damage to the pool structure remains local.
7.5 References for Section 7
[7.1.1]"OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978.
i l
l SIIADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 7-6 Report 111-981983
3 8 _
9 .
1
) 8
, c 9- _
e _
V 4/ i
(
s n 3 5
1 7
5 3
9 0
3 t
l H
t r
o p
e _
R _
2 2 1 _
0- 0- 0- _
e e e _
9 3 9 -
1 7 0 .
O 1 6 3 2
) 6 6 4 .
c 0 0 0 _
e + + +e
/
s e e N _
2 5 5 9 O _
, n 9 9 9 I K i
( 1 1 9 T A
S M E )
R I 2 O T n F I
i
/ 4 4 1 N C 2 c 0- 0- 0- I O e s e e e Y L - 9 9
9 9
0 R -
E b 0 A V A I
( 6 6 1 T E
T I 1
- C R P
7 e A P O 7 _
l e R -
b M c P 7 a I gf a T F ar z
) 9 9 2 S r u n 9 8 N DSAi O ( 2 9
2 2
4 I
A Y T R N t
h A g O M i e C M H T p X U o ) 5 0 E
S r n 6 0 8 T DHi ( 3 2 4 D E
D A
l i
t S h 0 g 0 i 0 0 e )
b 0 0 0 WWl (
5 1
5 1
1 1
y y k 9
l l c
b b a e m R p l e l y e s # e l us " u a n
,T FA F i
o t
a t
n p p m e o e v r o r
t n
E w o p D D I c
- p l p k c o
r l
ao e e c t t
h r a o D SD D R l I
3 _
8
. 9 .
1 1 1 1 0-
_ 0- 0- 8 _
e 9-
- r e e e _
u 0 0 0 i _
l i
0 8
0 0 H _
_ a 8 8 .
tr _
F 3 3 3 - o p
e _
R 4 4 2
- 0- 0- 0 n i c e e e t 7 7 1 a
i _
_ s 1 1 5 _
t r l a 7 7 8 .
S E 7 7 3 -
. 4 4 5
_ 0 0 0 _
+ N e +e + .
r e E O _
- u) l i 0
2 0
0 0
0 I
T a s A i
6 1 4 _
F (p 6 7 1 - M -
R O
_ F
_ d l
4 4 5 N S e 0 0 0 I _
+e +
E i
+ Y _
I s Y 0 e e R _
T s e t )
3 0
0 3
6 A R r si r s 1 5 0 T _
E S t i F (p 2 2 E _
2- P 1 - I _
R 7
O P e R O 8 .
l P 7 R - -
b 7 7 a L A
s u
0
+
0
+
0
+
+
P7 .
T cl e e e e S
I R iut 0 0 0 5 N sd) 6 6 6 0 I
E A aois 7 7 7 6 T l E M (p 2 2 T
A 2 3 N M O C
T X
y E i
t T
s) nf D ec 0 0 0 0 E 9 9 9 5 D (p 4 4 4 1 D A
l i
L S 4 4 0 0 0 3 3- 3 6-0 0 4 e 4 4 6 p 2 2 5 0 0
y A A A 0 T S S S 3 l
a e l l n m e e l
e o a e e e it t t t a N S S S n r
l a
s s
s s
s s t e t e
i r l e l e l e e r I n
t e i n n n c n c
. a a i
a i
a o e t
M S t
S t
S t
C l
I o
I
3 8
9
_ 1 8
44 44 4044 4 s
t 00 00 0300 0 r 33 - - 33 - - 36330
- - - - 3- 0 o p
00 00 0400 0 0 0 e e
p 44 22 44 22 4644 0 2522 3 4 0 2 3 R y AA AA A A A A '= A '=f, T SS SS SSSSf, S ii cc ii cc c c c c c iiiii c c i i t t t t t t t t s s aa s s aa s s s s st t s s t
t r l l l l aaaaa l l l l l l a a l l
a .o PP PP PPPPP P- P T i r
i v
oo - -
oo - - - -
ooooo o o E e a t t t t t t t t t t t G t h s s s s s s s s s s s a e aa aa a aaaa a a N R M B l l EE l l EE l l l l l EEEEE l
E E l
O A I T
T A D M N ee ee eeeee e e R A l a
l l bb l l bb l l l l l bbbbb l
b b l
O F
R r a a aa aa aaa a a u mm mm mmmmm mm N O t e e r r r r r r r r r r r I
T mpy ffoo ffoo ooooo fffff fof o Y C t ee e e eeeee e e R A T DD DD DDDDD DD A P T M n E I
3- I i
o r e m m R 7 F t s s d u u P p l l O 9 e O i r
l e
l e kci n t a
t a
l b S c C C olyd r r R -
a s l
l t t P7 N
l T O D e
Cn e g eg n
BCarS P e ne r
e S S t
n i t
ai l l aa gi t n e t N I
T t e r ei n l
n t t nLe i
L re I
e pi s si r A I
N g m mj o eo eerl c l o
c T o n r e sj dd aon I a l od ad ee eoo o N F T E CA BA PPBPC P C O E C D T L X A l a E _
R r T _
u -
oi c D U t e e d d d E T t t C my p s s aa i
g i
g i
g 2 D
A _
t l l i i i -
U T EP R R R 7 l R n l e i S -
T o b a
S i t T p n _
i r y y y i
c s
e l
b l
b l
b k d
t e
s D m m m c i e e a )
l r s) s) e Rb e o t sb sb s) sb r
a t n l l e l
A A l
c a
e 0 0 A 0 g 0 a 0 i s
e p m l e 50 l
e 0 l
e 0 r 0 ne e u 1 5 5 o1 m l u 1 u t p 1 E F( F( F(
1 S
1
( o l
a P
r n o
p l a
i t
t o i r a de ne r p p e m _
d k k p t .
o o o a e _
t aE v w r c r cl - M t n _
l o d ar r d aa r t d I up l p r e pr es k c e
t s o l
a e e ev n t
ee evd c or h e DOp e a l t
o PD S doc R t I
I
p l: 3 1.. i l
l I
.p -
l
)
IMPAGOR 1
. l L /
.b p
..;u.:.
2 .-5 r; -
- x" "-r.:.x.:- oxe l
-35
.s.. :.-x.
s.. . .....,,
m:s::sc#ma I
i ri 1, - . - - .
.J FIGURE 7-1 " SHALLOW DROP" SCENARIO
. +
Holtec international HI-981983 r
i
r; l +
l BASEPLATE , f l \ '
/
O O O O' 0 0 0 0 q
OOO IMPA00R IMPA[ TOR BASEPLATE ,
Mo oo O o o o-OOOO !
I p' - ~
- i FIGURE 7-2 " DEEP DROP" SCENARIO Holtec International Hi-981983 +
I
l l
l l
A*
'> \
l l
l I
i f +
Y .~
g .
3 X
FlOURE 7-3 ISOMETRIC VIEW OF THE " SHALLOW DROP" MODEL L' "1""
Holtec International HI-981983 I
_o
1 I
4.9943E+004 BWR HOLTEC RACK SHALLOW IMPACT 4.9943E+0 STEP 34 TIME = 8.4998854E-002 45066E+0 MAX _VONMISES 4.0189E+0 3.5313E+0 3.0436E+
2.5560E+
2.0683E+0 1.5806E+
1.0930E+0 l 6.0531E+0 1.1765E+0 1.1765E+003 ;
k vn Y X
FIGURE 7-4 VON MISES STRESS DISTRIBUTION IN STORAGE CELLS
""'".::1 Holtec International 111- 981983
STEP 34 TIME = 8.49988 1.1688E-002 Z COORDINATE DISPLA 1.1688E-0
-7.3217E-0
-1.4760E+0
-2.2199E+0,IN
-2.9637E+
-3.7076E+0
-4.4515E+ I
-5.1953E+0
-5.9392E+0 1
-6.6830E+0
-7.4269E+0
-7.4269E+00C Y
A Z > X s
FIGURE 7-5 PLAN VIEW OF POST-lMPACT GEOMETRY
.a ".1
~ T. "
Holtec International Hi-981983
I r- Ml M s' L- s s.
- -x ,
- s,
- %, s, i
r
- -x ,
^ s
! , -s s -s s !
^ ^
f- s, s.
L- x
, s.
-s ;
l- -s . ,
^ i
!.- ^ s. s ,
7 i -- s .-s E r
,s ,' ,s j k.27 Y M: *-
X l
l FIGURE 7-6 ISOMETRIC VIEW OF " DEEP DROP" MODEL l
7.:'"C.1 Holtec International HI-981983
1 2.9845E-003 STEP 21 TIME = 1.0499954E-002 2.9844E-0 Z COORDINATE DISPt.ACEMENT
-2.2296E-0
-4.4891E-0
-6.7486E- l
-9.0081 E-
-1.1268E+
-1.3527E+
.5787E+
l
-2. SE+00(
k
<5
- X o
FIGURE 7-7 MAXIMUM DISPLACEMENT OF RACK BASEPLATE
': ".". ~.L Holtec International Hi-981983
I
-. . ._-.2..__ ._ _.
_.k. '
ll ..
s.
- 1. w t . . . .. . :
FIGURE 7-8 ELEVATION VIEW OF " DEEP DROP" MODEL (ABOVE PEDESTAL) l l
l l
"7. "4"" "
Holtec International HI-981983
I l
5.8829E+002 8X8 BWR HOLTEC REGION 2 RACK DEEP CORN 5.8629E+0 STEP 18 TIME = 3.5998998E-003 4.9990E+0 SIGZZ(MID) 4.1351E+0 3.2711E+0 2.4072E+0 ,
1.5433E+0 l 6.7937E+0 E+0 002 k
X FIGURE 7-9 Z-DIR STRESS DISTRIBUTION IN CONCRETE
'LT "Z"."
Holtec International HI-981983
1 l
l l
8.0 SPENT FUEL POOL STRUCTURAL EVALUATION
\
8.1 Introduction {
l The Oyster Creek Nuclear Generating Station (OCNGC) Spent Fuel Pool (SFP) is a safety-related, seismic category I, reinforced concrete structure. At present, the OCNGS pool is equipped with 2,645 storage locations. This licensing report provides the technical basis to support GPUN's planned addition of four new high-density spent fuel racks in the open space of the pool to increase the total installed capacity to 3,035 storage locations. A plan view of the new rack configuration is shown in Figure 1.1. This chapter describes the analyses performed by GPUN's contractor, Holtec Intemational, to demonstrate the structural adequacy of the pool structure, as required by Section IV of the USNRC OT Position Paper [8.1.1].
1 The pool region is analyzed using the finite element method. Results for individual load 1 components are combined using factored load combinations mandated by SRP 3.8.4 (8.1.2] that incorporates the OT Position Paper. It is demonstrated that, for the bounding factored load combinations, the structural integrity of the pool is maintained when its slab is assumed to be fully loaded with 3,035 spent fuel assemblies, which represents the maximum pool's capacity after new rack installation. 1 The analysis summarized in this chapter parallels the previous pool structural adequacy evaluation carried out by GPU Nuclear to support the reracking of the OCNGS pool in the 1983-84 period. The only difference between the input quantities of the ca.1983 analysis (contained in the licensing report [3.1.4]) and the present work is the increase of the dead load (and the associated seismic adder) arising from the addition of 390 loaded storage cells. The effective increase represents 4.7% of the actual total dead load acting on the slab from the weight of the slab, the weight of the fuel transfer cask, the loaded racks, and the contained v'ater (see Table 8.3.1).
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 8-1 Holtec Report HI-981983
w The regions examined in the SFP are the floor slab and the undergirding support structure. Both moment and shear capacities are checked for concrete structural integrity. All structural capacity
- calculations are made using design formulas meeting the requirements of ACI -349 [8.1.3).
8.2 - Description of Pool Structure The spent fuel pool, which is elevated above ground, is located in the Reactor Building between -
.the 3rd floor (EL. 75'-3") and 5th floor (EL.119'-3"). Figure 8.2.1 shows a plan view of the
' portion of the reactor building where the spent fuel pool is located. As can be seen from Figure 8.2.1, the pool floor lies mostly between building column lines RC and RE and lines R5 and R6.
The inside (plan) dimensions of the pool are 27'-0" (North-South) by 39'-0" (East-West). The pool depth, which is measured from the top of the liner (80'-6") to the top of the pool curb (119'-
7"), is 39'-1". The interior surface is covered by a fully weld.ed, stainless steel liner.
The contents of the pool are supported by a two-way, reinforced concrete slab and underlying girders. The minimum thickness of the slab is 54 inches, excluding grout. The SFP walls to the north and to the east are 6 feet thick. The thickness of the west wall is reduced from 72 inches to l 54 inches above elevation 95'-3", where the new fuel storage pool is 'ocated. The south wall of I the pool is an integral part of the concrete reactor shield, and it has a minimum thickness of 6 l
feet. The walls are braced, from the outside, by several intermediate slabs.
{
l l A major contributor to the structural strenFth of the OCNGS slab is the lattice of deep girders, which are shown in Figure 8.2.1 by identihers RD and RE running along the north-south and R6 I mnning along the east-west direction. Section A-A from Figure 8.2.1 shown in Figure 8.2.2 illustrates the girder RD (N-S girder). Similarly,'Section E-E in Figure 8.2.3 provides a vertical cross section of the E-W girder (labeled "R6").
- As discussed later, the structural assessment of the OCNGS pool slab considers the above-i mentioned girders in one integrated finite element model.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 82 Holtec Report HI-981983
l 8.3 Annlicable Londinon
, The pool structural loading on the OCNGS pool- slab involves the following discrete components:
- a. Static (Dead) Loade
- 1. Dead weight of slab and girders.
1
- 2. Maximum dead weight of rack modules and fuel assemblies in the fully l implemented storage configuration,'as shown in Figure 1.1. '
- 3. Dead weight of a shipping cask (100 tons).
- 4. . The hydrostatic water pressure.
Table 8.3.1 provides a compilation of the dead loads on the OCNGS pool slab.
- b. Seismic Loade All spent fuel racks in the OCNGS pool are of the free-standing genre. During the postulated seismic events, the fuel assemblies rattle inside the storage locations, while the rack modules themselves respond to the seismic input as 3-D friction supported cantilevered structures. The Whole Pool Multi-Rack (WPMR) analysis
, methodology described in Chapter 6 of this licensing report provides the means to l compute the aggregate pedestal load on the pool slab (from all pedestals in the pool) as a function of time. The maximum value of the aggregate pedestal load, denoted by Psse and Pose for the SSE and OBE events, respectively, can be conservatively used as the equivalent static load on the slab. The seismic loads E and E' (corresponding to the OBE and SSE events) are, therefore, given by
[ E = Po,s - D 2 (8.3.1)
E' = Psse - D:
i L'
where D2is the total dead load from the fuel storage racks and the stored spent nuclear fuel. Numerical comparisons indicate that the gross pedestal loads Pass and Psse are bounded by their respective ZPA-based amplified dead loads, i.e.,
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 8-3 Holtec Report HI-981983
Poe, < ao.e D 2 (8.3.2)
P,s<asse 3 D2' where a, along with the appropriate subscript, denotes the ZPA for the vertical seismic component.
For conservatism, the right-hand side of the inequality (8.3.2) is used in lieu of their corresponding aggregate pedestal loads in equation (8.3.1).
- c. Thermal Loade Thermal loads are defined by the temperatures at the faces of the pool walls and slabs. For this analysis, two separate thermal loads are evaluated. The two cases result from consideration of the bounding summer and winter ambient temperatures at the site.
- d. Cask Droo Load The accidental drop of the cask onto the SFP slab is considered as an equivalent static load of1560 kips.
8.4 Analysis Model 8.4.1 Finite Element Model for Mechanical Loade The finite element model encompasses, in plan view, the entire SFP slab and its supporting beams delineated by column line R5 to the south, column line R7 to the north, column line RC to the east, and column line RF to the west. The preprocessing capabilities of the computer code STARDYNE [8.4.1] are used to develop the 3-D finite element model.
The STARDYNE finite element model contains 170 nodes,135 plate type finite elements,75 beam type finite-elements, and 43 concentrated masses, representing the pedestal loads of the fully loaded racks and cask. The plate type finite-element employed in the analysis has the capability to consider the shear deformation, which is essential for thick plates such as those representing the S ent Fuel Pool reinforced concrete slab. Figures 8.4.1 and 8.4.2 show the SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 8-4 Holtec Report HI-981983
1 l
i pictorial view of the numbered elements in the slab, girden, and peripheral walls. The corresponding node' designations are presented in Figure 8.4.3 for the pool slab and the girders.
Consistent with the existing design basis model [8.1.4], to simulate the interaction between the modeled region and the rest of the Reactor Building, a number of boundary restraints are imposed upon the described finite element model, The boundary conditions are shown in Figure 8.4.4. All nodes (Lthrough 16) located at the junction between the massive Reactor wall and the Spent Fnel Pool slab are fixed. The nodes (132,169, and 170) describing the end of the reinferced concrete beams at the exterior wall location, the node (113) representing the reinforced concrete column located at the intersection oflines RC and R6, and the nodes (54 and
- 63) corresponding to the location of the counterforts are all simply supported.
l l
8.4.2 Rgrmal Moment In order to obsain a conservative estimate of the thru-thickness moment in the concrete slab, due to a temperature gradient AT, across its two faces, the slab is assumed to be fixed against rotation along all four edges. The classical constrained elastic plate equation (using the cracked section 1 assumption permitted in ACI-349) is utilized to determine the thermal moment M,:
a, E, AT H' y'
12 (1- v,)
where:
l H= thickness of the slab (54 in.)
E, = effective Young's Modulus of the reinforced slab (with recognition for cracking ofconcrete in tension)
! v, = Poisson ratio (v, = 0.167) a, = concrete thermal expansion coefficient (a, = 5.5 x 104)
AT = difference between the top and bottom surface temperature of the slab SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 8-5 Holtec Report HI-981983 I
e _
y 8.5 Load Combinations The applicable load combinations germane to the OCNGS slab analysis are derived from reference (8.1.2] as follows:
- a. 1.4D + 1.9E
- b. 0.75 (1.4D + 1.4T,)
- c. 0.75 (1.4D + 1.4T,+ 1.9E)
- d. D + T, + E' + Y where:
D = total dead load on the slab E= OBE seismic load E' = SSE seismic load T, = thermal loading due to the temperature difference between the two opposing surfaces of the pool slab for summer or winter conditions.
Y= cask drop load The thermal loading actually consists of two cases. The temperature gradients in the summer and winter months are opposite of each other, thus leading to thermal moments which are opposite of each other in sign. The load cases (b) through (d) above, which include the thermal loading, T ,
actually comprise two distinct load combinations T_ and T% corresponding to summer and winter conditions, respectively.
. Therefore, the four (4) generic load combinations expand into eleven (11) load combinations, which are used in the numerical investigation. They are as follows:
Load Combination No.1 (LCl) = 1.4D + 1.9E Load Combination No. 2 (LC2) = 0.75 (1.4D + 1.7T_)
Load Combination No. 3 (LC3) = 0.75 (1.4D ' 1.7T.) ,
Load Combination No. 4 (LC4) = 0.75 (1.4D + 17 T_ + 1.9E) I Load Combination No. 5 (LCS) = 0.75 (1.4D + 1.7 T_ - 1.9E)
Load Combination No. 6 (LC6) = 0.75 (1.4D + 1.7 T% + 1.9E)
Load Combination No. 7 (LC7) = 0.75 (1.4D + 1.7 T% - 1.9E)
Load Combination No. 8 (LC8) = D + T_ + E' +Y SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 8-6 Holtec Report HI-981983 I.. ..
g Load Combination No. 9 (LC9) = D + T_ - E' + Y Load Combination No.10 (LC10) = D + T% + E' + Y Load Combination No.11 (LC11) = D + T.,,, - E' + Y 8.6 Innut Data for Mechanical and Thermal M The input data summarized in the following was used to perfonn the structural analyses:
Compressive strength of concrete, f,' = 3,000 psi j Yield strength of rebars, fy = 40,000 psi Top-to-bottom surface of the slab design basis temperature differential:
Summer: 110*F - 85'F = 25'F (poolside cooler) l Winter: 100*F - 40*F = 60'F (poolside warmer)
The above input data, along with the plant construction drawings, is utilized to compute the
.section bending and shear strengths and the thermal moments arising from the temperature gradients AT appropriate to T_ and T ,,,. !
l l - 8.7 Results i
l The shear forces and bending moments obtained, for all finite-elements describing the slab and j its supporting beams, for all eleven (11) load combinations are compared with the cross-sectional capacities of the specific structural element. The safety factors are calculated as the ratio of the
! reinforced concrete cross-sectional capacity to the corresponding calculated cross-sectional load.
The minimum calculated safety factors for bending and shear are summarized in Table 8.7.1. All calculated safety factors are greater than 1.0 for the postulated load combinations. Also given in Table 8.7.1 are the load cases leading to the minimum safety factors.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 8-7 Holtec Report HI-981983 L
i j 8.8 Pool I iner 1
The pool liner in~ the OCNGS pool is not a safety-related component. The "non-safety related"
, designation for the pool liner arises from the fact that its failure would not cause a rapid lowering of the water level in the fuel pool. Nevertheless, it is demonstrated that the rack pedestal and bearing pad design do not produce a state of overstress in the liner, leading to liner failure from
. tearing or cyclic fatigue during seismic events. !
Like all fuel pool liners, the OCNGS pool liner is subjected to the weight of water (plus any l seismic load) and lateral plus vertical forces from individual spent fuel rack support pedestals. I 1
The liner is a stainless steel plate resting on a concrete substrate and loaded laterally by the water ;
l- and the loaded spent fuel racks. Calculations have been performed to show that the liner will not l tear or rupture under all loading conditions in the pool and that the liner can withstand one SSE and twenty OBE seismic events without fatigue failure. The cumulative damage factor under one '
SSE and twenty OBE events is 7.35 x 10", which is well below the acceptance limit of 1.0. A strength evaluation of the liner plate section has also been performed. The maximum stress in the liner and the weld seams during an SSE seismic event has been computed and found to be less than the allowable stress limits. It is concluded that the liner in the OCNGS pool will maintain its integrity during and after the postulated seismic events. ;
8.9 Closure The analyses summarized in this chapter demonstrate that the safety factors in the pool slab and the undergirding concrete beams exceed 1.0 for all postulated load combinations. The integrity of the pool liner under the in-plane loadings during the postulated seismic events is also demonstrated. Inasmuch as the total increase in the pool slab dead load under the proposed storage capacity addition is less than 4.7%, the analyses summarized in this chapter essentially l reconfirm the veracity of the existing design basis established during the ca.1983 rerack of the l OCNGS pool.
l l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 88 Holtec Report HI-981983
8.10 References for Section 8
[8.1.1) OT Position for Review and Acceptance of Spent Fuel Handling Applications, by B.K. Grimes, USNRC, Washington, D.C., April 14,1978.
[8.1.2] NUREG-0800, SRP-3.8.4, Rev. I., July 1981.
[8.1.3] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit Michigan.
[8.1.4] GPU Nuclear, Licensing Report on High-Density Spent Fuel Racks for Oyster Creek Nuclear Generating Station, NRC Docket No. 50-219, August 1983.
[8.4.1] STARDYNE User's Manual, Research Engineers, Inc., Rev. 4.4, July 1996.
1 l
l l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 8-9 Holtec Report HI-981983
l i ,
1 I
1 Table 8.3.1 MPLICABLE LOADINGS ON THE OCNGS POOL SLAB Item Category Value ,
1
- 1. Self-weight of the Dead Load (Di ) 1,530 kips reinforced concrete mass
- 2. Total weight of the racks and Dead Load, (D2) 2,441 kips stored fuel
- 3. Weight of fuel transfer Dead Load,(D3 ) 200 kips cask
- 4. Gross downward weight of Dead Load, (D4 ) 2,628 kips contained pool water Sum Total of Dead Loads, Drn, 6,799 kips SHADED TEXT CONTAINS PROPRIETARY INFORMAllON l
Holtec International 8-10 Holtee Report HI-981983 l :
l L. l
Table 8.7.1 CONTROLLING SAFETY FACTORS AND LOAD CASES Structural Bending Shear Element N S Direction - E-W Direction N-S Direction E-W Direction Beam RD 2.06 (LC8) -
2.96 (LCl) -
Beam RE 1.19 (LC8) -
2.92 (LC1) -
Beam R6 -
1.06 (LCl) -
1.43 (LC8)
Slab 1.45 (LC6) 1.71 (LC6) 1.11 (LCl) 1.09 (LC2)
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 8-11 Holtec Report HI-981983 I
- . ~ . -
A m T-1 1
= _8, h I m w a w 1 f I l E M
1 I l
lWil l I
'W) i
)
l 1
I l I I i i i i 1 1 I i
_____ ____ ,, , ,,a _ _ _ _ _ _ _ _ _,_,_ m _ _ ,
__________a i i,
,._________i__l___,,
4 6 4, g i i i gg i l I I i Il,,tg . ,g g
Ii. 3 I I g
I I
g gI
\ _ _ _ _ _ _ _ _ _ _j l sl i___g , ___ _y__fl____7 vi' 3r -------7',r'-'---*-- - T '( -W- L'-- j u i 1 , ,
r_______ ___. '
I I i
I 4
I I 1 i l I
' I I I 1 I i I I I . ,17 g g 3 ,I l l l 41 1 g I
g g i I h I g I 'g l i 1 1 .4+ l g T
i l i l i f I i g O.
l t i I I I i 1 I I i l i i l L7 i i l i e I7 I I
= _ . __ &..J ...-_ J 8 L____ . _g ' ' '
_ ___ _ ' ' - - x
'- _. ,_ l_
Q l
= -
_ _ _ _ -- J y__
s a ,-
I J i ka
- . I s
\
l I i
- W j
/
j i \
\
l I
h
/ \ l I g 61
/
g\ l I \l I gi e i
._ _____c I ____ q d.
E
-____f .--___J O
\ g Il
=, \
- ll
\ I g
\ d g
\ u.a i
% I 1 /i I s , I e 1 I #
1 I
i ,' i i i
ri >
1 g
' 3 g
)
--,r------
i i
-- 7 r'------- --- J i i f t t 0
,I r... "] , I I I v i , i i l j I l
I
[ g LJ
- i i %
W
.a
3 8
9
- 1 8
9 i
H d
Wr R e
t u E O D R
M' I
a
, ' i ,
G
,l i!
S-N
_ L gt A
_ g .
R
_ T N
E g_b ,
C g _
' H 7- G
. * ,L 3
1 jt A U O
I i -
_ A
, l a
N O R H
l I
_ T C
E T
S N L
_ - O I
L OE _ T E PN UEU
_ t s C FGL We e r E TAT S NRS _ t si r d
EON PTT aG
_ E 2 SSS _
_ 2 0
1 9
/ _
8 G
7 _ I L _ F E j _
J 3
8 9
1 8
9 I _
~
H
%,,a ' ' .
n R
_ E D
R I
_ G f_
W -
E
- b I
_ L M _
A I _
_ R c , T N
E
, l , , C H
_ '-T
, E E
N O
I T
G U
O R
C H
~
a ,
_ 3 E
_ 1 S T
_ N cl , ,
O I
T C
E S
, l 0'
1 9'- _ 3 7
L 2
E n F
I y 8 cl G
I a F
% i i
4'
.l l i a 4- g r l ll
3 _
8 91 8
9 I
H B
A L
S F
O s
- N O
I T
A Z
I T
E R
C S
I D
T N
E M
E 9 "
L E
E T
I m
N I
F 1
4 8
GI F
X Iz 1 1 1 VLC i
7 -
3 8
9 1
8 9
i H
S R
E D
R I
6 3 G
1 1
1 1 D N
A 5 2 S 1
1 1
1 L L
A 4
1 1
1 W
1 1 F 3
4 1
4 0 0 8 7 6 5 4 131 3 3 1 3 3 3 4 2 3
1 3
0 3
9 2
O -
1 8 N
1 1 1 1 1 1 1 1 6
215_
5 1 8 2 9 8 7 1
5094 9_ 5_
1 7
2 O -
7 6 5 H3 02 0 18 B I
T 2 7 61 1
y0 0 1
l 1' 1' 1
0 I 0 G 96 1
5~13 1 I 1 1 I 1 1 1 9 A
1 9 2 9
~
2 1
5 2 i 4 3 2 9 8 l 1
8 1
t 4-1 9
2 8
4 2 R -
1 1 C
7 4 0 1 3 2
SI 9
1 8 1 D
6 4 9 8
0 2 2 T 1 8 1 N 5- 8 9 1 E 4
1 8 2 1 M 4 7 7_ 0 E
4 8 8 2 L _
1 7_ 1 E .
E T
I N
I F _
2 _
4 8
X G
I F
' 2 1 1 1 VLC
0 9 7 6 1
1, S
8 R
6 7
6 E 1 1 D
R I
6 5 G 6 6 1 1 D N
A 4
6 6 3 2 6 6 1 0 9 8 7 6 61 5 15 5 5 5 4 3 5 5 5 2
5 1
5 0
5 9
4 B 1 1 1 1 1 1 1 1 1 1 1 1 1 1 A
8 L 1 0 14 9 S 3 3 2 M yj!l 1 1 1 2 R 1
1 ~
s O
F 6
9 S N
0 8
OI T
4 A
6 .
N G
I 8
4 S
E 2
3 D
E 6
1
- D O
N 3
4 8
G I
F X
y 2 1 1 C
- ! b 3
8 9
1 8
9 i
H A3 b3 a
S _
N O
I
_.. T I
D _
N O
C
,) I '
]
Y
\
a a R A _
D N
U O _
B 4
4 8 _
a G
I _
+ s 4
F _
- 3 _
1 1 2 _
1 1 1 1 1 1 1 1 1 1 1 1 1 Y
I 1 1 VLC
n.
I 9.0 RADIOLDGICAL EVALUATION 9.1 Fuel Handling Accident -
The potential radiological consequences at the Oyster Creek Nuclear Generating Station exclusion area boundary (EAB) of a fuel handling accident in the spent fuel pool have been determined.
9.1.1 - Assumntions and Source Term Calculations Evaluations of the accident were based on fuel of 4.6 wt% initial enrichment burned to 40,000 Mwd /mtU. The reactor was assumed to have been operating at 1930 Mw thermal power with 560 fuel assemblies in the core prior to shutdown. The fuel handling accident was assumed to result in
- the release of the gaseous fission products contained in the fuel / cladding gaps of all the fuel rods in
' a peak-power,8x8 fuel assembly. Gap inventories of fission products available for release were estimated using the release fractions identified in Regulatory Guide 1.25 [9.1.1] except for Iodine-131, for which the release fraction is increased 20% in accordance with NUREG/CR-5009 [9.1.2].
Cooling time for the failed fuel prior to the accident was 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
The gaseous fission products that have the greatest impacts on the off-site doses following short fuel-cooling periods are the short-lived nuclides ofiodine and xenon, which reach saturation inventories during in-core operation. These inventories depend primarily on the fuel specific power over the few months immediately preceding reactor shutdown. In the high'est power assemblies, the specific power and hence the inventory of iodine and xenon will be proportional to the peaking factor (taken as 1.50, per Reg Guide 1.25).
At the cooling time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> used in the Oyster Creek calculations, more than 90 percent of the thyroid dose comes from Iodine-131, while nearly 90 percent of the whole-body dose comes from Xenon-133. Though these iodine and xenon isotopes are the major contributors to off-site doses, the contributions from other radionuclides are calculated and included in the overall dose values.
SHADED TEXT CONiAINS PROPRIETARY INFORMATION Holtec Intemational 9-1 Report HI-981983
The present evaluation uses values for atmospheric diffusion factor (x/Q) and for filter efficiencies that have been specified previously. Core specific inventories (Curies per metric ton of uranium) of fission products were estimated with the ORIGEN code [9.1.3] based upon parameters stated earlier (Initial enrichment of 4.6 wt% U, bumup of 40,000 Mwd /mtU, and a cooling time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />), along with a specific power of 20.0 kw/kgU. The msults of the ORIGEN calculations for isotopes that am considered as possible contributors to'the thyroid, whole-body, and skin doses are given in Table 9-1,'while Table 9-2 lists pertinent data for the isotopes of interest. Data and assumptions used in the dose calculations are given in Table 9-3.
The equation given on the following page was taken from Reg Guide 1.25 and was used to calculate the thyroid dose (D,in rads) from the inhalation of radiciodine. Values for the terms in the equation are given in Table 9-1, Table 9-2, and Table 9-3.
F, I t F P B R. (x/Q) i DF, DF, F, = fraction of fuel rod iodine inventory in gap R, = dose conversion factor (rads per space curie)
I, = core iodine radionuclide inventory at time of x/Q= atmospheric diffusion factor the accident (curies) (see per cubic meter)
F= fraction of core damaged so as to release DF,= effective iodine decon. factor for iodine in the rod gap pool water P= core peaking factor DF,= effective iodine decon factor for filters B= breathmg rate (cubic meters per second)
The equation given below was used to calculate the extemal whole-body dose from gamma radiation in the cloud of radionuclides released in the fuel-handling accident. The equation contains several of the terms defined above.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 9-2 . Report HI-981983
I Dose, = E 0.25 - (X/Q) F P G, E,...
1 In this expmssion, G,is the gap inventory of the gaseous radionuclides of krypton and xenon, while the E term is the average energy per disintegration of each radionuclide (in Mev per disintegration, as given in Table 9-2). The equation assumes the noble gas decontamination factors in water and the charcoal filters are 1.0. - The gap inventories of radioiodine make negligible contributions to the whole body dose, D, , because of the large decontamination factors applicable to the iodines.
9.1.2 Results The doses at the Oyster Creek EAB from the specified fuel handling accident are tabulated below.
The doses am based on the release of all gaseous fission product activity in the gaps of all fuel rods in a highest-power assembly.
Thyroid dose, rad = 0.487 Whole-body dose, rem = 0.1%
Skin dose, rem = 0.725 These potential doses are well within the exposure guideline values of 10 CFR Part 100, paragraph
' 11. As defined in Standard Review Plan 15.7.4, Radiological Consequences of Fuel Handling Accidents, "well within" means 25 percent or less of the 10CFR100 guidelines, or values of 75 rad for thyroid doses and 6.25 rem for whole-body doses. The potential doses at Oyster Creek from the conservative scenario pmsented here meet the criteria for "well within."
9.2 Sohd Radwaste The necessity for resin mplacement is determined primarily by the requirement for water clarity, and the msin is normally changed about thme to four times a year. No significant incmase in the volume
. of solid radioactive wastes is expected with the expanded storage capacity. During rack installation SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 9-3 Report HI-981983
( _
U operations, a very small amount of additional resins may be generated by the pool cleanup system on a one-time basis.
l 9.3 Gaseous Releases l
Gaseous releases fmm the fuel storage area are combined with other plant exhausts. Normally, the contribution from the fuel storage area is negligible compared to the other releases and no significant increases in the overall releases are expected as a result of the expanded storage capacity.
9.4 Personnel Exoosures During normal operations, personnel working in the fuel storage area are exposed to radiation from the spent fuel pool. Operating experience has shown that the area radiation dose rates, which originate primarily from radionuclides in the pool water, are generally 1.0 to 2.0 mrem /hr, with a few areas.such as the pool bridge showing slightly higher dose rates.
Radiation levels in zones surrounding the pool are not expected to be affected significantly. Existing shielding around the fuel (water, stainless steel pool liner, and concrete wall) provides more than adequate protection, despite the slightly closer approach of the new racks to the wall of the pool.
Operating experience has shown that there have been negligible concentrations of airborne radioactivity, and no increases are expected as a result of the expanded storage capacity. '
. No increase in radiation exposure to operating personnel is expected; therefore, neither the current health physics program nor the area monitoring system needs to be . modified.
9.5 Anticioated Exoosure During Re-racking All rack installation operations will utilize detailed procedures prepared with full consideration of ALARA principles. Similar operations have been performed in numerous facilities in the past, and there is every reason to believe that rack installation can be safely and efficiently accomplished at
- Oyster Creek, with minimum radiation exposure to personnel.
1
- SHADED TEXT CONTAINS PROPRIETARY INFORMATION
! IIoltec International 94 Report HI-981983
f' Total occupational exposure for rack installation operations is estimated to be between 0.7 and 1.4 person-rem, as indicated in Table 9-4. While individual task efforts and exposures may differ from those in Table 9-4, the total is believed to be a reasonable estimate for planning purposes. Divers I will not be used, but the person-rem burden includes a figure for their possible exposure if there is an emergent need for them.
l The existing radiation protection program at Oyster Creek is adequate for the rack installation operations. Personnel will wear protective clothing, activities will be governed by a Radiation Work l l
Permit, and personnel monitoring equipment will be issued to each individual in accordance with approved plant procedures.
Work, personnel traffic, and the movement of equipment will be monitomd and controlled to assure that exposures are maintained ALARA.
9.6 References for Section 9
[9.1.1] Regulatory Guide 1.25 (AEC Safety Guide 25), " Assumptions Used For Evaluating The Potential Radiological Consequences Of A Fuel Handling Accident In The Fuel Handling And Storage Facility For Boiling And Pmssurized Water Reactors", March 23,1972.
[9.1.2] C. E. Beyer, et al., " Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors", NUREG/CR-5009, Pacific Northwest Laboratory, Febmary,1988.
[9.1.3] SAS2H-ORIGEN-S/ARP, in " Scale 4.3-Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation",
IRTREG-CR-0200, Rev. 5, September 1995.
l l
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 9-5 Report HI-981983 I
l Table 91 RESULTS OF ORIGEN CALCULATIONS FOR RADIONUCLIDES OF IODINE, KRYMON, AND XENON AT 100-HOURS COOLING TIME (Extremely small values are shown as zeroes)
Radionuclide Curies ner mtU I-131 3.865 x 105 I-132 3.240 x 105 I-133 7.544 x 105 I-134 0.000 x 108 I-135 2.719 x 10' I
Kr-85 1.035 x 10' Kr-85m 2.380 x 104 Kr-87 0.000 x 10 Kr-88 0.000 x 10 Xe-131m 6.433 x 10' Xe-133 7.544 x 105 Xe-133m 1.415 x 10' Xe-135 1.468 x 10 3 Xe-135m 4.439 x 10 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 9-6 Report HI-981983
1 Table 9-2 l RADIONUCLIDE PROPERTIES USED IN THE FUEL HANDLING ACCIDENT ANALYSIS Dose l
Conversion, Radionuclide Rads / Curie Ey (Mev/ dis.)
Iodine-131 1,48 x 10' lodine-132 5.35 x 10' -----
Iodine-133 4.00 x 105 -----
Iodine-134 2.50 x 10' -----
Iodine-135 1.24 x 105 -----
Krypton-85 -----
0.002 i Krypton-85m -----
0.157 Krypton-87 -----
0.792 I'
Krypton-88 1.954 4
Xenon-131m -----
0.020 Xenon-133 -----
0.047 Xenon-133m -----
0.040 1 l
Xenon-135 ----
0.249 Xenon-135m -----
0.428 l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 9-7 Repon HI-981983
1 ,
Table 9.3 DATA AND ASSUMPTIONS FOR THE EVALUATION OF THE FUEL HANDLING ACCIDENT Core power level, Mw(t) 1930
]
Fuel enrichment, wt% U 4.6 Fuel bumup, Mwd /mt U 40,000 Specific power, kw/kg U 20.0 i
Fuel cooling time, hrs 100 Power peaking factor 1.50 l
No. of failed fuel rods All in 1 assembly
)
' Core inventory released to gap, %
Iodine-131 i 12 Otheriodines 10 Krypton-85 30 Xenon-133 10 Other xenons 10 lodine composition, %
Elemental 99.75 Organic 0.25 Pool decontamination factors Elemental iodine 133 Organic iodine 1 Noble gases 1 )
Filter decontamination factors ;
Elementaliodine 100 '
Organic iodine 100 l Noble gases 1 Atmospheric diffusion factor x/Q),sec/mv 7.60 x 10 4 i Breathing rate, m'/sec 3.47 x 10' SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 9-8 Report HI-981983
T .
1 i
l Table 9-4 i
l PRELIMINARY ESTIMATE OF PERSON-REM EXPOSURES l FOR RACK INSTALLATION l
l Estimated
! Number of Person-Rem Operation Personnel Hours Exoosure' l
Clean and vacuum pool 2 10 0.1 to 0.2 Determine swing bolt locations 2 5 0.2 to 0.4 j Install new racks 5 30 0.4 to 0.8 Total Exposure, person-rem 0.7 to 1.4 Assumes minimum dose rate of 2-1/2 mrem /hr (expected) to a maximum of 5 mrem /hr except for pool vacuuming operations, which assume 4 to 8 mrem /hr, and possible diving operations, which assume 20 to 40 mrem /hr.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Internationa! 9-9 Report HI-981983
'10.0 BORAL SURVRIH ANCE PROGRAM y
l 10.1 Purpose .
Boral", the neutron absorbing material incorporated in the spent fuel storap rack design to assist in controlling system reactivity, consists of finely divided particles of bomn carbide (B.C) uniformly distributed in type 1100 aluminum powder, clad in type 1100 aluminuni~and pressed and sintered in a hot-rolling process. Tests simulating the radiation, thennal and chemical environment of the spent fuel pool have demonstrated the stability and chemical inertness of Boral (References [10.1.1) through [10.1.3]). The accumulated dose to the Boral over the expected rack lifetime is estimated to be about 3 x 10' to 1 x 10" rads depending upon how the racks are used and the number of full-core off-loads that are necessary.
Based upon the accelerated test programs, Boral is considered a satisfactory material for reac-tivity control in spent fuel storage racks and is fully expected to fulfill its design function over the lifetime of the racks. Nevertheless, it is prudent to establish a surveillance program to monitor the integrity and performance of Boral on a continuing basis and to assure that slow, long-term synergistic effects, if any, do not become significant. Furthermore, the April 14,1978 USNRC letter to all power reactor licensees (Reference [10.1.4]), specifies that
" Methods for verification of long-term material stability and mechanical integrity of special poison materials utilized for neutron absorption should include actual tests."
The purpose of the surveillance program is to characterize certain properties of the Boral with the objective of providing data necessary to assess the capability of the Boral panels in the racks to continue to perform their intended function. The surveillance program is also capable of d-tecting the onset of any significant degradation with ample time to take such corrective action as may be necessary.
In response to the need for a comprehensive Boral surveillance program to assure that the suberiticality requirements of the stored fuel array are safely maintained, a surveillance program SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 10-1 Report HI-981983
l has been developed incorporating certain basic tests and acceptance criteria. The Boral surveill-
~
l' ance program depends primarily on representative coupon samples to monitor performance of the absorber material without disrupting the integrity of the storage system. The principal parameters to be measured are the thickness (to monitor for swelling) and boron content.
,10.2 Counon Surveillance Pronram l 10.2.1 Coupon Dea 11ption The coupon measurement program includes coupons suspended on a mounting (called a " tree"),
placed in a designated cell, and surrounded by spent fuel. Coupons will be removed from the l anay on a prescribed schedule and certain physical and chemical properties measured from l which the stability and integrity of the Boral in the storage cells may be inferred.
Each surveillance coupon will be approximately 4 inches wide and 6 inches long. The coupon surveillance program will use a total of 10 test coupons with each coupon mounted in a stainless i steel jacket, simulating as nearly as possible, the actual in-service geometry, physical mounting, materials, and flow conditions of the Boral in the storage racks. The jacket (of the same alloy used in the manufacture of the racks) will be closed by screws or clamps to allow easy opening 1
with minimum possibility of mechanical damage to the Boral specimen inside. In mounting the j coupons on the tree, the coupons will be positioned axially within the central 8 feet of the fuel l zone where the gamma flux is expected to be reasonably uniform.
l Each coupon will be carefully pre-characterized prior to insertion in the pool to provide initial reference values for comparison with measurements made after irradiation. The surveillance l coupons will be pre-characterized for weight, length, width and thickness. In addition, two coupons (which need not be jacketed) will be preserved as archive samples for comparison with subsequent test coupon measurements. Wet chemical analyses of samples from the same lot of Boral will be available from the vendor for corpparison. I l
l l SHADED TEXT CONTAINS PROPRIETARY INFORMATION 1 Holtec Intemational 10-2 Report HI 981983 I L l
! 10.2.2 ~ hveill-a Connon Tantine Schaan1e l
i l The coupon tree is surrounded by freshly discharged fuel assemblies at each of the first five refuelings following installation of the racks to assure that the coupons will have experienced a i slightly higher radiation dose than the Boral in the racks. Beginning with the fifth load of spent
)
fuel, the fuel assemblies will remain in place for the remaining lifetime of the racks. The l coupon management schedule is shown in Table 10-1.
i- At the time of the first fuel off-load following installation of the coupon tree, the (8) storage cells I 1
l surrounding the tree shall be loaded with freshly-discharged fuel assemblies that had been among the higher specific power assemblies in the core. Shortly before the second reload, the coupon tree is removed and a coupon removed for evaluation. The coupon tree is then re-installed and, I
at reload, again surrounded by freshly discharged fuel assemblies. This procedure is continued for the third, fourth, and fifth off-loading of spent fuel (except that a coupon is not pulled at the fourth refueling). From the fifth cycle on, the fuel assemblies in the (8) surrounding cells
- remain in place.
l l Evaluation of the coupons removed will provide information of the effects of the radiation, l
\
\
l themial and chemical environment of the pool, and, by inference, comparable information on the Boral panels in the racks. Over the duration of the coupon testing program, the coupons will
! have accumulated more radiation dose than the expected lifetime dose for normal storage cells.
i l
l Coupons that have not been destructively analyzed by wet-chemical processes may optionally be l returned to the storage pool and re-mounted on the tree. They will then be available for subsequent investigation of defects, should any be found.
1 i
l _10.2.3 Measurement Program l
The coupon measurement program is intended to monitor changes in physical properties of the Boral absorber material by performing the following measurements on a scheduled basis:
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 10-3 Report HI-981983
i l
l
- a. Visual observation and photography,
- b. Neutron attenuation, :
l
- c. Dimensional measurements (length, width and thickness),
- d. Weight and specific gravity, and
- e. Wet-chemical analysis (a process wherein the aluminum in Boral is chemically dissolved in an acid solution leaving Boron carbide precipitate which can be dried and weighed to determine the B4 C content in the coupon).
The most significant measurements are thickness (to monitor for swelling) and neutron attenuation' (to confirm the concentration of Boron-10 in the absorber material). In the event that loss of boron is observed or suspected, the data may be augmented by wet-chemical analysis I (a destmetive gravimetric technique for total boron only).
10.2.4 Surveillance Connon Accentance Criteria Of the measurements to be performed on the Boral surveillance coupons, the most important are:
(1) the neutron attenuation measurements (to verify the continued presence of the boron) and (2) the thickness measurement (as a monitor of potential swelling). Acceptance criteria for these measurements are as follows: I
- a. A decrease of no more than 5% in Boron-10 content, as determined by neutron i attenuation, is acceptable. (This is tantamount to a requirement for no loss in L boron within the accuracy of the measurement.)
l b. An increase in thickness at any point should not exceed 10% of the initial thickness at that point.
Neutron attenuation measurements are a precise instrumental method of chemical analysis for Boron
-10 content using a non-destructive technique in which the percentage of thermal neutrons trans-mitted through the panel is measured and compared with pre-determined calibration data. Boron-10 is the nuclide of principal interest since it is the isotope responsible for neutron absorption in the Boral panel.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION l
Holtec Iniemational 10-4 Report HI-981983
l l
l l
Changes in excess of either of these two criteria requires investigation and engineering evaluation which may include early retrieval and measurement of one or more of the remaining coupons to provide corroborative evidence that the indicated change (s) is real. If the deviation is determined to be real, an engineering evaluation shall be performed to identify further testing or any corrective action that may be necessary.
The remaining measurement parameters serve a supporting role and should be examined for early indications of the potential onset of Boral degradation that would suggest a need for further attention and possibly a change in measurement schedule. These include (1) visual or photographic evidence of unusual surface pitting, corrosion or edge deterioration, or (2) unac-countable weight loss in excess of the ac .iurement accuracy. l 10.3 In-Service Innnection (Blackness Tests)
In-service inspection involves directly testing the Boral panels in the storage racks by neutron logging' (sometimes called " blackness testing"). This technique is able to detect areas of l significant boron loss or the existence of gaps in the Boral, but cannot determine other physical properties such as those measured in the coupon program.
! In the event that the surveillance coupon program shows a confinned indication of degradation, blackness testing may be one of the techniques employed to investigate the extent of degradation, l if any, in the racks. Blackness testing is a method of comparing the data found in the pool with a known sample. There is no acceptance criteria per se, but a comparison of readings taken of the 1
Boral in the racks with those taken of known sample allows one to determine where there may be deficiencies in the Boral panels in the spent fuel pool storage racks.
l l
l l ' Neutron logging, is a derivative of well-logging methods successfully used in the oil industry for l many years.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION l Holtec Intemational 10-5 Report HI-981983
4 I
l 10.4 References for Section 10 l .
[10.1.1] " Spent Fuel Storage Module Corrosion Report," Brooks & Perkins Report 554, June 1,1977.
l
[10.1.2] " Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in i PWR Storage Pools," Brooks & Perkins Report 578, July 7,1978.
l l [10.1.3] "Boral Neutron Absorbing / Shielding Material - Product Performance l l
~ Report," Brooks & Perkins Report 624, July 20,1982.
I
[10.1.4] USNRC Letter to All Power Reactor Licensees transmitting the "OT !
l Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978.
I l
1 l
l l
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 10-6 Report HI-981983
)
[.
/
r l
Table 10-1 COUPON MEASUREMENT SCHEDULE Coupon Refuelingt After Rerack i
1 1st" I 2 2nd" q 3 3rd" i
4 5th" 5 8th 6 1Ith 7 14th 8 20th j 9 25th l 10 30th Remove coupons for evaluation within 1 or 2 months before the next refueling.
Place freshly discharged fuelin the 8 surrounding cells at the beginning of the 1st,2nd,3rd,4th, and 5th refueling cycles after completion of reracking.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 10-7 Report HI-981983
i l
11.0 INSTAT T ATION 11.1 introdnetion The construction phase of the Oyster Creek spent fuel pool rack installation will be executed by GPU Nuclear personnel. All construction work at Oyster Creek will be performed in compliance with NUREG-0612 and site-specific procedures.
I Crane and fuel bndge operators are to be adequately trained in the operation ofload handling machines per the requirements of ANSI /ASME B30.2-19% and the plant's specific training program.
1 The lifting device designed for handling and installation of the new racks at Oyster Creek is in compliance with the provisions of ANSI N14.6-1978 and NUREG-0612, including compliance 1 with the primary stress criteria, load testing at a multiplier of maximum working load, and nondestructive examination of critical welds.
An intensive surveillance and inspection program shall be maintained throughout the rack installation phase of the project. A set ofinspection points which have been proven, in numerous previous rack installation campaigns, to eliminate any incidence of rework or erroneous installation will be implemented.
Holtec Intemational and GPU Nuclear have developed a complete set of operating procedures which cover the entire gamut of operations pertaining to the rack installation effort. Similar
- procedurt.s have been utilized and successfully implemented by Holtec International on previous rack installation projects. These procedures assure that ALARA practices are followed and provide detailed requirements to assure equipment, personnel, and plant safety. The following is
- a list of procedures that will be used to implement the rack installation phase of the project.
l l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 11-1 Report HI-981983
i I
- a. Installation / Handling Procedure:
l t
This procedure provides direction for the handling / installation of the new high-density racks.
The procedure delineates the steps necessary to receive new high-density racks on site, to unload and upright the racks, to stage the racks prior to installation, and to install the racks. The procedure also provides for the installation of new rack bearing pads, adjustment of the new rack pedestals and performance of the as-built field survey.
- b. Receipt Inspection Procedure:
This procedure delineates the steps necessary to perform a thorough receipt inspection of the new racks after their arrival on site. The receipt inspection includes dimensional measurements, cleanliness inspection, and visual weld examination.
- c. Cleaning Procedure:
This procedure provides for the cleaning of the new racks, ifit is required, in order to meet the requirements of ANSI N45.2.1, Level B. Permissible cleaning agents, methods and limitations on materials to be employed are provided.
- d. Pre-Installation Drag Test Procedure:
This procedure stipulates the requirements for performing a functional test on the new racks prior to installation into the spent fuel pool. The procedure provides direction for inserting and withdrawing a " dummy" fuel assembly into designated cell locations, and establishes an acceptance criterion in terms of maximum kinetic drag force during withdrawal.
- e. Post-Installation Drag Test Procedure:
This procedure stipulates the requirements for performing a functional test on new racks ,
following installation into the spent fuel pool. The procedure will provide direction for inserting and withdrawing a " dummy" fuel assembly into designated cell locations, and establishes an l
I acceptance criterion in terms of maximum kinetic drag force.
j
- f. ALARA Procedure:
Consistent with both Holtec Intemational's and the Oyster Creek plant's ALARA Programs, this !
procedure provides details to minimize the total man-rem received during the rack installation l project, by accounting for time, distance, and shielding. Additionally, a pre-job checklist is i l established in order to mitigate the potential for overexposure, l
~
- g. Liner Inspection Procedure: 1 In the event that a visual inspection of any submerged portion of the spent fuel pool liner is deemed necessary, this procedure describes the method to perform such an inspection, which ,
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 11-2 Report HI-981983
l l
l includes the use of an underwater camera, and describes the requirements for documenting any observat. ions.
- h. Leak Detection Procedure:
This procedure describes the method to test the spent fuel pool liner for potential leakage using a vacuum box. This procedure may be applied to any suspect area of the pool liner.
- i. Underwater Welding Procedure:
In the event of a positive leak test result, an underwater welding procedure may be implemented which provides for the placement of a stainless steel repair patch over the area in question. The procedure contains appropriate qualification records documenting relevant variables, parameters, and limiting conditions. The weld procedure is qualified in accordance with AWS D3.6-93, Specification for Underwater Welding or may be qualified to an alternate code accepted by GPU Nuclear and Holtec Intemational.
j j. Job Site Storage Procedure:
l l
This procedure establishes the requirements for safely storing the new racks on site, in the event l
)
l that long-term site storage is necessary. This procedure provides environmental restrictions, temperature limits, cleanliness requirements, and packaging requirements.
I1.2 Rack Arranoement )
The existing Oyster Creek rack arrangement consists of 10 racks, which represent 2,645 cell locations. The proposed rerack is a supplemental expansion that will add four new racks to the spent fuel pool. The racks are to be installed along the north wall adjacent to the Cask Drop ;
l Protection System (CDPS). The four new racks provide a total of 390 additional storage locations. This rerack does not require any of the existing racks to be removed or displaced. A l schematic depicting the spent fuel pool after completion of the rerack can be seen in Figure 1-1. )
i
)
11.3 Pool Survev and inenection ;
A pool inspection shall be performed to determine if any items attached to the liner wall or floor will interfere with' the placement of the new racks or prevent usage of any cell locations subsequent to installation.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 11-3 Report HI-981983
W ;
In the event that protrusions are found which would pose any interference to the installation process l appropriate actions will be taken to ensure safe rack installation.
i l
i 11.4 Pool Coolina and Purification '
11.4.1 Pool Cooline The pool cooling system shall be operated in order to maintain the pool water temperature at an acceptable level. It is anticipated that specific activities, such as bearing pad elevation i i
measurements, may require the temporary shutdown of the Spent Fuel Pool cooling system. At no time, however, will pool cooling be terminated in a manner or for a duration that would create a violation of the Oyster Creek Technical Specification.
I1.4.2 Purification The existing spent fuel pool filtration system shall be operational in order to maintain pool clarity. Additionally, an undenvater vacuum system shall be used as necessary to supplement fuel pool purification. The vacuum system may be employed to remove extraneous debris, reduce general contamination levels prior to diving operations, and to assist in the restoration of pool clarity following any pressure washing operations.
I1.5 Fuel Shufflina As new high-density racks are installed into the pool, it is anticipated that fuel shuffles will be performed in independent phases in order to transfer irradiated assemblies from existing racks into Holtec racks. .
Fuel shuffle operctions shall be conducted in accordance with station procedures and in a manner consistent with the rack installation sequence. Final shuffle operations shall not be conducted during new rack installation.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International _ 11-4 Report HI-981983
l I
11.6 Installation of New Racks The new high-density racks, which are supplied by Holtec International, shall be delivered in the horizontal position. Each new rack shall be removed from the shipping trailer using a suitably rated crane, while maintaining the horizontal configuration, and placed upon an upender and secured. Using two independent overhead hooks, or a single overhead hook and a spreader beam, the module shall be uprighted into the vertical position.
I 1
The new rack lifting device shall be installed into the rack and each lift rod successively engaged.
Thereafter, the rack shall be transported to a pre-leveled surface where the appropriate quality control receipt inspection shall be performed.
In preparing the spent fuel pool for the rack installation, the pool floor shall be inspected and any debris that may inhibit the installation of bearing pads shall be removed. After the bearing pads l are positioned on the pool floor, elevation measurements shall then be taken at each pad location. l
! The rack pedestals shall be adjusted in accordance with the bearing pad elevation measurements j in order to achieve levelness when the racks are installed.
- The new racks shall be lifted using the 100-ton overhead crane. A temporary hoist with a suitably rated capacity may be attached to the overhead crane for installation activities in order to avoid submerging the main hook and causing contamination. For rack movements along the pool floor, the height of the rack above the liner shall not exceed six inches, except where floor l
projections obstruct the path. Once the rack has reached its final position it shall be carefully lowered onto its bearing pad. A proposed rack installation sequence for Oyster Creek is shown in Figures Il-1 through 11-4.
Elevation readings shall be taken to confirm that the module is level and as-built rack-to-rack and rack-to-wall offsets shall be recorded. The lifting device shall be disengaged and removed from the fuel pool under Radiation Protection direction. A post-installation drag test may be l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 11-5 Report HI 981983
C L
i-performed using an inspection gage to ensure that no cell location poses excessive resistance to the insertion or withdrawal of a bundle.
I1.7 Safety. Radiatinn Protection. and ALARA Methoda
~
11.7.1 Safety During the rack installation phase of the pmject, personnel safety is of paramount importance, outweighing all other concerns. All work shall be carried out in strict compliance with approved procedures.
11.7.2 Radiation Protection Radiation Protection shall provide necessary coverage in order to provide radiological protection and monitor dose rates. The Radiation Protection department shall prepare Radiation Work Permits (RWPs) that will instruct the project personnel in the areas of protective clothing, L
general dose rates, contamination levels, and dosimetry requirements.
In addition, no activity within the radiologically controlled area shall be carried out without the knowledge and approval of Radiation Protection. Radiation Protection shall also monitor items removed from the pool or provide for the use of alarming dosimetry and supply direction for the ,
proper storage of radioactive material. I 11.7.3 ALARA The' key factors in maintaining project dose As Low As Reasonably Achievable (ALARA) are time, distance, and shielding. These factors are addressed by utilizing many mechanisms with respect to project planning and execution.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 11-6 Report HI-981983
(-
l t
Time Each member of the project team will be properly trained and will be provided appropriate i
education and understanding of critical evolutions. Additionally, daily pre-job briefings will be employed to acquaint each team member with the scope of work to be performed and the proper means of executing such tasks. Such pre-planning devices reduce worker time within the radiologically controlled area and, therefore, decrease pmject dose.
i Distance Remote tooling such as lift fixtures, pneumatic grippers, a support leveling device and a lift rod i
disengagement device have been developed to execute numerous activities from the pool surface, where dose rates are relatively low. By maximizing the distance between radioactive sources and project personnel, project dose is reduced.
Shielding During the course of the rack installation, the water in the spent fuel pool provides primary shielding. The amount of water between an individual at the surface and an irradiated fuel assembly is an essential shield that reduces dose. Additionally, other shielding may be employed to mitigate dose when work is performed around high dose rate sources.
I1.8 Radwaste Material Control Radioactive waste generated from the rack installation effort may include vacuum filter bags, miscellaneous tooling, and pmtective clothing.
Vacuum filter bags may be removed from the pool and stored as appropriate in a suitable container in order to maintain low dose rates.
l-L SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International Il-7 Report HI 981983
I' i Contaminated tooling shall be properly stored per Radiation Protection direction throughout the l
project. At project completion, an effort will be made to decontaminate tooling to the most practical extent possible.
l i
l l
l l
l l
i 1
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 11-8 Report HI-981983 i
)
1, a l L
I V
h t
I T
- I me I F
D I
t l
A A 1 1
M T
S N
Fl )
1 R
V W
l I 1 F
. = g g s I
=
P I
g T N E _
g g E I
l I
T E
E I
G S.
. g@
g g
T S
W D N l P.
= g ( _
= ll; f
e (b. , l
= _
l E _
l E
C. C F
- .ll e, N
lI
-[i l E
U sf Q
f E
L i[ o K _ _
E C S i!/ .
c R
A _
M G
N iI 0 _
. 0 I
I T
S N
@ E E@ O l, I I
- X _
E
_ N _
T i
A ll.
]'
T O
!_ 0 o
L L
._ " : . - "l A iI o [!I i T qe : : : : , 1 l l l .
S
- 7'_
- N i)
- " _ _ _ I
- '_, iI
, ' re : : : , i l E
l l -
K C
A C K
C R A A
R
_ G N
R K
C G@
N I
T I
S I
K A
R T S B E l_ E g I
I E
E I
T S
R I
I C
E R
l a an n" i
-l ,
,,I :: E T
S
_ _ = Y
_ _ = 0 E E C C 1 A A R R -
1 G
N I
T S
I E
I T
S I
@ =
=
@ 1 X
E X
E E R
_ _ = U
_ _ =
~
G
!i I
N e F
I fi ii a ' ,- l : l l l
e a
m t
u a
_ 2
_ 3 i 8 l
_ 9 G I
Bi 9 N
- P l
i a u H
[
f t
I T
A l : : , Is i L
L A
T 6 N Fi )
i l
f W -
f I
i\ ~
_ =
2 l D
I F _ M
[
._ P N
E U
A T
[
I E
E I
C ~ S.
. g @
T S
M D N _ = (
I I
P. _ _
- t s!; l (b. , :
=
I E X
E
~ C. C F
Il 'l ::
N
_ :: I l<,, iI E
k U M ~ iili Q E
l o
i[/ E
~
E L
l E E -
7i
!i C
A C A - S l!
R R O
E@ N ri E@
_ g I T
S I
T O I S I I
_ I E X E
T A
{l_ l._ g
!i r
L _
_ o L _
_ - o ,, " :l l
. . il
" A _
g ib ;.
O -I i T .
) k' N - _ - S
(
, d I l
i: !
' q':,,i l l N
1 ,,
4 : _ - 1 I
_. , ' ] g yI W O v
- = -
,l I
K K _
-g i' , il ;':,,f C C _
E C
! ' R l l ;
A
- A = -
R E
C A
=
R E@
I
=
E@
I S
I T
K R T S E I
E E
I X
E J__ E nBU I
T : = -
R S =
I I
C E _
R
,l ll '! ga n" - ,, :: l. E T
S _
_ = = _ Y
_ = =
O I I K C
A C
A C
A 2
R R R -
1 E =
E = _
_ I T
S I
= I T
S I
@ =
E@
I T
S I
1 X
E X
E I
E E R
_ =
=
=
=
U G
!i I
N l t it -
F i' : .
- l m
a m
m a
r
_ a t
)
1 1
5 m
I N
F N
[
m G I
L D
I I
T :
I
- A A ,i L
L L A
I S
L rl ) G r
N D I
i _ = E 3 l H
F _
C A N
=
P I I R _
N E H
[
I T
C E
I G S.
. g E@
I T
S I
T S
l W
D N I _
- = E ( _
I P. _
T
{b. e i II_ =
I X
E _
E C. C F _
4" :
N _
. l.,,, ::li l E
0 U
t E
, l Q
E L
s
, N K = -
E
,j C
A = C K
A - S R
I o R N
G 0' r!
I 0 o N
I T
S
@ E@
I T
S O
I I X =
I f
E
= E I -
T A
0 o! L
~ -
- _ o L O - o -
o #" : l - . " A I
)
(
i !:
n o l T h ,' N S a I O -
o o o
n
-i y$
c' r
' it 4 i i : t d
=
' : , il l i N
I
,t' ii K
K rt I ' i I 1 : 0 '
C C Rs A
l l itKC
=
A R
- C N
R K
C A
=
E@
I
= I T
S I
K R T S
X E E I
E E@
I T
S I
I
=
=
I E =
=
R C
E R
l n8 1 = 4 - .,.. :: E T
S
_ = = _ Y
_ = =
0 K E K C
A C
A C
A 3
R R R _
_ C 1 E =
E =
N
_ I T
S I
= I T
S I
@ = I T
S I
@ 1 I
E I
E X
E E R
=
=
= U
=
G
!I I _
N g 8 F .
i i ' ,' : .'
l e _
a -
m m
a n
a
- 3 4, l
8 i 9
1 E
- 2 9
N P
l
. i I
. N - H K
[ l t
_ I T U 1
- : : 4 A .
L A
T S
N L
r )
1 B
N
)
1 I
' _ 4 J F
F
[
, = ,
G T
N
= , _ P I I
I E
[
,g l
I T , l i
i E
E .
T N R
I G _ S.
S D N I P.
(
- lI ,
' yb. g : iIl_ .
I E C. C U_F N 1I ,j< s < l E
]
U M _
gi l
Q E
E -
I O E L
i/ I a =
m
~
S oj(
i j
7i o
u =
o ~
o _ o ;._ '
ri ' m N m@
o o i L
_ _ o U i
a =
m@
m ~
O I
T d' - = ~
A i ! .
ji
_ o o _.ll._ o U L _
o o0 _
. - O L o
. _ o _ - O . -l A
!_ o o _ g ) - ,[: l " l _
i . i l.
( -
i T _
r' , \ - _ - S _
i_
-i
- ' _ , l-l
, q,l . , ; l l N _
i i J.
1 I
. o o _. g -
_U - _ - = ~
o o0 - yI V _
K 1 lI r
m C I ' I l
' , :ml , ;= ,
l
, , 9l l ,
A .
= o , ~ R _
G
= ,@
= ,
s @ ~
K H 1 s m E l
E m E I
T S
- =
=
~
R I
X
- ~
C E
R I " "ll gN I a ] - l .., ll E T
S
- = = _ Y
- = =
0 a " m 4 n " o _ -
1 E =
" =
- I T
s
=
" @ =
m@ 1 i
n " m _
E R
=
=
= U
=
G I
_ N l ' ' i i , ' : : ;
F j
B
- I E
I t
R F
[
R P
C E
T L
H
12.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 12.1 Introduction Article V of the USNRC OT Position paper [12.1.1] suggests the submittal of a cost / benefit analysis for the chosen fuel storage capacity enhancement method. This section abstracts the analyses and evaluations made by GPU Nuclear before selecting reracking as the most viable )
alternative.
1 I
12.2 Imperative for Rerackina The specific need to increase the existing storage capacity of the Oyster Creek spent fuel pool is based on the continually increasing inventory in the pool, the prudent desire to restore full core offload capability, and a lack of viable economic alternatives. 1 Reference is made to Tables 1-1 of Section I wherein the historical and projected fuel discharges for the Oyster Creek spent fuel pool are tabulated. The table shows that the plant has already lost the capability to offload one full core (560 fuel assemblies) to the spent fuel pool. Reracking is the only alternative to restore full-core offload capability by the next operating cycle.
12.3 Appraisal of Altemative Options GPU Nuclear has determined that storage expansion is by far the most viable option for the Oyster Creek plant in comparison to other alternatives.
The key considerations in evaluating the alternative options are provided below.
i
- a. Timeliness: restores full-core offload capability as soon as possible
- b. Safety: minimize the number of fuel handling steps
- c. Economy: minimize total installed and O&M cost
- d. Security: protection from potential saboteurs, natural phenomena r
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 12-1 Report HI-981983 L
l
- e. Non-intrusiveness: minimize required modification to existing systems
- f. Maturity: extent ofindustry experience with the technology
- g. ALARA: minimize cumulative dose due to handling of fuel Spent fuel pool expansion was found by GPU Nuclear to be the most attractive option with respect to each of the foregoing criteria. An overview of the alternative technologies considered by GPU Nuclear is provided in the following.
Rod Consolidation Rod consolidation has been shown to be a feasible technology. Rod consolidation involves disassembly of spent ruel, followed by the storage of the fuel rods from two assemblies into the volume of one and the disposal of the fuel assembly skeleton outside of the pool (this is considered a 2:1 compaction ratio). The rods are stored in a stainless steel can that has the outer dimensions of a fuel assembly. The can is stored in the spent fuel racks. The top of the can has an end fixture that matches up with the spent fuel handling tool. This permits moving the cans in an easy fashion.
Rod consolidation pilot projects in the past have consisted of underwater tooling that is 5 manipulated by an overhead crane and operated by a maintenance worker. This is a very slow and repetitive process.
The industry experience with rod consolidation has been mixed thus far. The principal advantages of this technology are: the ability to modularize, moderate cost, no need of additional land and no additional required surveillance. The disadvantages are: the release of fission gases due to rod breakage, the potential for increased fuel cladding corrosion due to scraping of the protective oxide layer, the potential interference of the (prolonged) consolidation activity with ongoing plant operation, the increased dead weight and floor loading, and the lack of sufficient industry experience.
SilADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec International 12-2 Repon lil-9819f,3 t
l On-Site Cask Storace Dry cask storage is a method of storing spent nuclear fuel in a high capacity container. The cask provides radiation shielding and passive heat dissipation. Typical capacities for BWR fuel range from 44 to 68 assemblies that have been removed from the reactor for at least five years. The casks, once loaded, are then stored outdoors on a seismically qualified concrete pad. The pad will have to be located away from the secured boundary of the site because of site limitations.
The storage location will be required to have a high level of security that includes frequent tours, reliable lighting, intruder detection, and continuous visual monitoring.
The casks, as presently licensed, are limited to 20-year storage service life. Once the 20 years has expired the cask manufacturer or the utility must recertify the cask or the utility must remove the spent fuel from the container. In the interim, the U.S. DOE has embraced the concept of multi-purpose canister (MPC), which may make existing licensed cask designs less desirable. Work is also continuing by several companies to provide an MPC system that will be capable oflong term storage, transport, and final disposal in a repository.
Finally, facilities must be provided to vacuum dry the cask, back fill it with helium, perform leak checks, remachine the gasket surfaces ifleaks persist, and assemble the cask on-site.
Presently, no MPC cask has been certified for both storage and transponation. Because of the continued uncertainty in the government's policy, the capital investment to use a dry storage system is considered to be an inferior alternative for Oyster Creek at this time.
SilADED TEXT CONTAINS PROPRIETARY INFORMATION llottec International 12-3 Report 111-981983
( .. . .
Modular Vault Drv Storace Vault storage consists of storing spent fuel in shielded stainless steel cylinders in a horizontal configuration in a reinforced concrete vault. The concrete vault provides radiation shielding and missile protection. It must be designed to withstand the postulated seismic loadings for the site.
At the present time, no MPC technology based vault system has been licensed for fuel transport.
The high cost and uncertainty make this option less prudent. The storage modules installed at OCNGS are not MPC compatible and can not be rendered operable in a comparable time frame.
Horizontal Silo Storace A variation of the horizontal vault storage technology is more aptly referred to as " horizontal silo" storage. This technology suffers from the same drawbacks that other dry cask technologies have, namely,
- a. No fuel with cladding defects can be placed in the silo.
- b. Concern regarding long term integrity of the fuel at elevated temperature.
- c. Potential for eventual repackaging at the site.
- d. Potential for fuel handling accidents.
l
- e. Relatively high cumulative dose to personnel in effecting fuel transfer (compared i
to rcracking). l
- f. Compatibility of reactor / fuel building handling crane with fuel transfer hardware.
- g. Potential incompatibility w-ith DOE shipment for eventual off-site shipment. ,
I
- h. Potential for sabotage.
l I
l SilADED TEXT CONTAINS PROPRIETARY INFORMATIO.N llottec International 12-4 Report 111-981983
New Fuel Pool Constructing and licensing a new fuel pool is not a practical alternative for Oyster Creek t.ince such an effort may take up to 10 years. Moreover, the cost of this option is prohibitively high.
An estimate of relative costs in 1998 dollars for the aforementioned options is provided in the following:
Reracking: $2.5 million Horizontal Silo (NUHOMS): $10 to 15 million Rod consolidation (rental): $25 million Metal cask: $15 to 20 million Modular vault: $56 million New fuel pool: $150 million GF Nuclear's estimate of comparative costs of various options is consistent with other published industry data [12.3.1,12.3.2].
To summarize, there are no acceptable alternatives to providing full-core offload capacity in the next operating cycle for Oyster Creek Nuclear Generating Station. First, there are no commercial independent spent fuel storage facilities operating in the U.S. Second, the adoption of the l Nuclear Waste Policy Act (NWPA) created a de facto throw-away nuclear fuel cycle. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocessing represents an added cost for the nuclear fuel cycle which already includes the l
NWPA Nuclear Ware Fund fees. In any event, there are no domestic reprocessing facilities.
l l
12.4 Co:.t Estimate 1
1 The proposed construction contemplates the reracking of the Oyster Creek spent fuel pool using free-standing, high density, poisoned spent fuel racks. The engineering and design work is i completed for the project. This rerack provides enough additional storage capacity to restore full core of00ad capability until the Fall of 2000.
SIIADED TEXT CONTAINS PROPRIETARY INFORMATION 11oltec International 12-5 Report 111-981983 j l
)
f The total capital cost is estimated to be approximately $2.5 million as detailed below. 1 Engineering, design, project management: $1 million Rack fabrication: $1 million Rack installation: $ 0.5 million l As described in the preceding section, several alternatives were considered prior to proceeding j with new rack installation, which is not the only technical option available to increase on site storage capacity. New rack installation does, however, provide a definite cost advantage over other technologies.
12.5 Resource Commitment The expansion of the Oyster Creek spent fuel pool capacity is expected to require the following I primary resources.
Stainless steel: 18 tons t
l Boral neutron absorber: 6 tons (5 tons Boron Carbide, I ton aluminum) l The requirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than 0.001%). Although the fraction of world production of Boron Carbide l
required for the fabrication is somewhat higher than that of stainless steel or aluminum, it is I unlikely that the commitment of Boron Carbide to this project will affect other related activities.
Experience has shown that the production of Boron Carbide is highly variable and depends upon need and can easily be expanded to accommodate worldwide needs.
12.6 Environmental Considerations This rerack is not expected to increase the maximum bulk pool temperature above the previously licensed value. Therefore, the cooling water demand on Barnaget Bay and the water vapor emission to the environment should remain unchanged.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION llottec International 12-6 Report 111-981983
12.7 References for Section 12 (12.1.1] USNRC Letter to All Power Reactor Licensees transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling
. Applications," April 14,1978, and Addendum dated January 18,1979. ,
I (12.3.1] Electric Power Research Institute, Report No. NF-3580, May 1984.
[12.3.2] " Spent Fuel Storage Options: A Critical Appraisal," Power Generation )
Technology, Sterling Publishers, pp. 137-140, U.K. (November 1990).
J i
l i
l I
SilADED TEXT CONTAINS PROPRIETARY INFORMATION lloitec International 12-7 Report 111-981983 l
V l l
I i
i ILNfLOSURE 3 !
l Licensing Report for Storage Capacity Expansion of OCNGS Spent Fuel Pool Holtec Report 111-981983, Revision 4 Proprietary l
I I
l l