ML20154J602
| ML20154J602 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/31/1988 |
| From: | Charnley J, Elliott P, Lambert P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20154J579 | List: |
| References | |
| 23A5841, 23A5841-R, 23A5841-R00, NUDOCS 8809230081 | |
| Download: ML20154J602 (35) | |
Text
i 00 71 2]A5841 REVISION O CLASS I JULY 1988 23A5841 REV. O SUPPLEMENTAL RELOAD LICENSTM SGMITTAL FOR LA SALLE COUNTY STATION UNIT 2 RELOAD 2, CYCLE 3 Prepared by h dJ P. A. Lamb'ert Fuel Licensing Verified by: C e. ?b P. E. Elliott Fuel Licensing
/
s
/
Approved by:
S. Charnfey anager Fuel cens s g
$$$92gggS[
9 0., jy P
G GENwlearEnerpy
(
11S Conw Am tj2 \
- . s+aur. ca ssus
- 23A5841 REV. 0 IMPORTANT NOTICE RECARDING
()(} 5k) l CONTENTS OF THIS R1 PORT !
I t
PLEAS 2 READ CAREFULLY This report was prepared by General Electric solely for Commonwealth Edison Company (CECO) for Ceco's use with the United States Nuclear f 1
Regulatory Commission (USNRC) for amending Ceco's operating license of the .
La Salle County Station, Unit 2. The information contained in this report is believed by General Electric to be an accurate and true reprnsentation ,
of the facts known, obtained or provided to General Electric at the time l this report sas prepared. l l
The only undertakings of the General Electric Company respecting I
information in this document are contained in the contract between 0:m=envaalth Edison Company and General Electric Company for nuclear fuel and related services for the nuclear system for La Salle 2, and nothing
- ntained in this document shall be construed as changing said contract.
The use of this information except as defined by said contract, or for any
- urpose other than that for which it is intended is not authorized; and -
vith respect to any such unauthori:ed use, neither General Electric Company nor any of the contributors to this document makes any representation or -
warranty (express or implied) as to the completeness accuracy or useful- 1 ness of the information contained in this document or that such use of such j information may not infringe privately owned rightst nor do they assume any responsibility for liability or damage of any kind which may result from (
such use of such information. (
1 I
l l
l 3/4 l u
?
23A5841 REV. 0 ACKNUWLEDGEMENTS The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Submittal, were performed in the Nuclear Fuel and Engineering Services Department by F. T. Bolger.
4 5/6
l
. ,. 23A5861 REV. 0 l M
00 74 l
- 1. . PLANT-UNIQUE ITD4S (1,0)* . ;
i Transient Analysis Basis: Appendix A $
GETAB and Transient Analysis Initial Conditions: Appendix B
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0) 1 !
i Fuel Tvoo Cycle Loaded Number ,
a d j Irradiated ;
6 I
I 8CRB176 1 40 l 80RB219 1 260 224 !
! BP8CRB299L 2 j !
l ;
i New l-3C320C 3 96 1 BC300D 3 144 1
4 .
Total 764 l 4
)
- I i .
l l
. 3. REFERENCE CORE LOADING PATTERN (3.3.1) i
! Nominal previous cycle core average exposure at end of cycle: 17990 mwd /MT j
l Minimum previous cycle core sverage exposure at end of cycle from cold shutdown c*onsiderations: 17734 mwd /MT j Assumed reload cycle core average exposure at end of cycle: 19377 M'id/Mr i
- Core loading pattern Figure 1
)
a i *( ) Refers to area of discussion in General Electric Standard Application '
l for Reactor Fuel. NEDE-24011-P A 8 (dated May 1986): a letter "S" i preceding the number refers to the United States Supplement. .
s l 1 l 1
{
j 7.
h.
1 . ,
23A5841 REV. O !
i 00 75 !
l
! 4. CALCULATED CORE EFYECTIVE MULTIPLICATION AND CONTROL SYSTEM '
WCRTH- NO VOIDS J DEG. C ( 3. 3.2.1.1 AND 3. 3.2.1. 2 )
- 't i ;
Beginning of Cycle, k-effective ;
[
- Uncontrolled 1.123 l 1
Tully Controlled 0.960 l Strongest Control Rod out 0.985 4
R. Maximum Increase in Cold Core Reactivity with l Exposure into Cycle, Delta k 0.005 !
i a
- 5. STANDBY LIQUID CONTROL SYSTU4 SHVTDOVN CAPABILITY (3.3.2.1.3)
) l l
Shutdown Margin (Delta k) l (20 den.C. Xenon Free) j l m i I
! 660 0.037 (
1 i i i 1 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2) i
'.'alues normally reported in this section are REDY inputs. There were !
(
.o transisnts analyzed using REDY. t I
{ l 1 7. RELOAD-t'NIQUE GETA3 TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS I (S.2 .2)
Exposure: SOC 3 TO EOC3 1 '
i Fuel Peakina Factors . R- Bundle Bundle Flow Initial I Oesian M M M Factor Power (MWT) (1000 lb/hr) MCPR i
- GE8x8E3* 1.20 1.57 1.40 1.051 6.681 114.9 1,21
)
i I
1 I
l *All fuel types are bounded by G18x8EB.
I i
1 8 4 .
. o 23A5841 REV. 0 l .
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2) .
Transient Recategorization: No i
Recirculation Pump Trip: Yws l
No Rod Withdrawal Limiter:
Thermal Power Monitor: Yes Improved Scran Times No Exposure Dependent Limits: No Exposure Points Analyzed: 1 i
- 9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3) 1 Single-Loop Operation
- Yes l
i Load Line Limit: No l i
Extended Load Line Limit: Yes Increased Core Flow Yes a i 4 Flow Point Analyzed: 105% ,
l Teodwater Temperature Reduction: Yes j
No l i ARTS Program t Maximum Extended Operating Domain No 1
l l
1 i
l t
i i
l 4
- 1
. '. 23A5861 RIV. 0 00 77
- 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
Methods Used: GEMINI 't Delta CPR .
Flux Q/A Transient (% NBR) (% NBR) GE8x8EB* Ilaure Exposure Ranges BOC3 to EOC3 :
Load Rejection Without 383 115 0.14 2
[
j Bypass l
0.11 ** i
Exposure Range BOC3 to ECC3 Extended Load Line Limit Load Rejection Without 328 113 0.12 4 3
Bypass i
Feedwater Controller 211 109 0.08 5 Failure 4
Exposure Range Extended ECC with Increased Core Flow 4 Load Rejection Without 414 115 0.15 6 Bypass Feedwater Controller 260 112 0.12 7 Failure Exposure Ranges Extended EOC with Increased Core Flow and Final
- Feedwater Temperature Reduction Lead Rejection Without 374 114 0.14 8 Bypass Feedwater Controller 276 117 0.16 9 Failure
- All fuel types are bounded by GE8x8E3.
- See Appendix A.
10 i
,e
. , 23A5861 REV. 0 00 78
- 11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRtHENT FAILURE) i TRANSIENT
SUMMARY
(S.2.2.1) ,
i Limiting Rod Pattern: Figure 10 Rod Block Rod Position Delta CPR HLHGR (kW/ft)
Readina % GE8x8EB* GE8x8EB*
l (Feet Withdrawn) l
\
l 104 4.0 0.15 18.84 105 4.5 0.18 18.84 l 0.19 18.84 l 106 5.0 :
! 107 6.5 0.22 18.84 108 12.0 0.23 18.84 l 0.23 18.84 ,
l 109 12.0 110 12.0 0.23 18.84 l Set Point Selected: 106 l 12. CYCLE MCPR_ VALUES (S.2.2) l l Son-Pressurization Events j i
l Exposure Range BOC3 to EOC3 l
GE8x8EB*
l Loss of Feedwater Heating 1.18 I Fuel Loading Error Rod Withdrawal Error 1.26 4
Pressurization Events Exposure Range BOC3 to EOC3 l l
1 I
Option A Ootion B '
l GE8x8EBS GE8x8E8*
Load Rejection Without Bypass 1.26 1.22
. Feedwater Controller Failure 1.20 1.18 I
I 1
- All fuel types are bounded by GE8x8E3. <
- Tuel Loading Error not applicable for C Lattice plants. l 11 l
A _
33A5861 REV. 0
()l) 1Pf)
Fressurization Events Exposure Range: BOC3 to EOC3 Extended Lead Line Limit Option A Option B GE8x8EB* GE8x8EB* -
Load Rejection Without Pypass 1.24 1.20 1.10 1.17 Feedwater Controller Failure Exposure Range Extendad IOC uith Increased Core Flow Option A Oprion B GE8x8EB* GE8x8EE*
l Load Rejection Without Bypass 1.27 1.23 .
Feedwater Controller Failure 1.22 1.20 l
I Exposure Range Extended EOC with Increased Core Flow and Final Feedvater Temperature Reduction 1
l Option A Option 3
~
GZ8xdEBu CE8x8EB*
i Load Rejection Without Bypass -
1.26 1.22
- Feedwater Controller Failure 1.26 1.24 J
i 1 13. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3)
I Steam Line Vessel Pressure Pressure Plant transient M (esia) Response MSIV Closure 1227 1264 Figure 11 l (Flux Scram) l 0
]
- All fuel types are bounded by GE8x8EB.
]
- 12
'. 23A5841 REV. 0
()() k3()
- 14. LOADING ERROR RESULTS (S.2.5.4)
Not applicable for C-Lattice plants.
- 15. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
Centrol Rod Drop Accident Analysis is not required for banked position withdrawal sequence plants. NRC approval is documented in NEDE-24011-P-A-8-US, May 1986.
- 16. STABILITY ANALYSIS RESULTS (S.2.4)
GE SIL-380 recommendations have been included in the pla~t .
procedures and/or Technical Specifications: therefore, no e.
analysis is required. NRC approval for deletion of a cycle sp(
stability analysis is documented in NEDE-24011-P-A-8-US.
- 17. '0SS-OF-COOLANT ACCIDINT RESULTS (S.2.5.2)
- 0CA Method Used:
. SATER/GESTR-LCCA
.a Salla County Station Units 1 and 2. "SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31510P, 2ecember 1987 (as imended).
Technical Specification MAPLEGR Limits Puel Type BC320C (GE8x8EB)
MAPLEGR (KV/ f t)
Average Planar Exposure (CVd/ST) Most Limiting Least Limiting 0.0 11.6 11.6 3.0 -- 12.2 4.0 12.2 12.4 6.0 12.6 12.8 8.0 12.9 13.1 10.0 13.1 --
12.5 13.0 13.0 15.0 12.7 12.7 35.0 10.2 10.2 45.0 8.5 8.6 50.0 6.1 6.1 13
I 23A5841 REV. 0
()() b33.
- 17. LOSS-OF-COOLANT ACCIDENT RESULTS (S.2.5.2)(continued) ;
Technical Specification MAPLHGR Limits ,
Tuel Type BC300D (CE8x8EB)
Average Planar Exposure 4 (GWd/ST) MAPLHGR (KW/f t) c 9
i 0.0 11.8 i
- 2.0 12.3 3.0 12.6 l 4.0 12.9 ;
5.0 13.2 10.0 13.5 13.2 !
- 15.0 I
) 35.0 10.7 j 45.0 8.7 ,
1 50.0 6.6 j
i e
iI :
1 i
l i
t I
l 1 ,
1
, I f
1 !
4 <
, t
$ l 1
i i I i
i l
1 14 1 l
. . 23A5841 REV. 0 -
00 82 6: @ @itI TA C CT Q3Ij l s smF@Q@mF@rcem m .
b ll eaM M M M M M M M M M Moe l
- - MMMMMMMMMMMMMo ;
- MMMMMMMMMMMMMMM i
- MMMMMMMMMMMMMMM
- s-I,a r,I rie ra na sa se e s s a am se ca me sI rim
- i - 2 f l 6 00 f l G f 30 6 1 d l d l iM f l f D C I f l f 30
- 4- ria m,m a flID
- -Ilf3 r cia s aflFC c a s c FII es@ s m s m e r ta s I c,cT c aFIfif30I'00 EII3 F3
- i- FI FI FI FI fl 18 FI TQ TI E'I I I ET TI FI fi !
- i
- 4
,IIBIS rrcem EQIJTiQI3 Iwf@DTrrC EIDI y IJ QQI5ISI h rIcser25 ;
- ~
IILI15ESTJI5IEIIIgT3EIT3IE :
fd CGC 08 DQ- D8 CQ ft CI0: F0 CI '
'i ' C C j
- ' ' ' ' ' ; i 5MMMMMMMT lllllll l l l l l l l l l l l l l l l t 357 9111315171521232527
- 9313335373941434547495152555753 TUEL TYPE A = 8CRB176 D = BP8CRB299L 8 = 8CRB219 E = BC300D <
C = 8CRB219 F = BC320C Tigure 1. Reference Core Loading Pattern t$
1 2
23A5841 REV. 0 l 00 83 1 1
4 J l 1
I 1 hD teena FLLk I 1'Ill0. eng!g alsgtPg])
2 3aig spractt.<st it.ws *Ft" #8 bit 8L0s 3 (08E !%CT fGv ,,,,,
! ((g {f 8" 9?
I g g ,, g ]
i I
1 i
j .
! $ 18 3. 3 210.8 a Y N 4 s 1~ w _ .
18.3 384,8
' _N q
Lu 1 i ...
t, s
(
- 2. s 4. 8 1.8
- 8. 8
_/
- 2. 3 1.0 f.t I f!*( ttt: N il flat (!!: M il l
] i i
. .g.g. t ;i.;w.s tr. lgs.pa r t i i <c:: startiv;te
( *O S I'cp*'.testaCftvtff
. Sttape g as
' I' s: ti s t'* !8;?T.I?!?"'4"' .: A i" 'I'iU '!"
l
. \ / \ _
f i t. i
'+ 88 - -
~ q%/
! l l
)
84 8 U['VN Q M a i k...f~., . ;
i l _
~
W el .1 -
j
)
+ t 8 8. 8 3,6
... 3. . ... ... ... ... ... ...
l f!M titt M ll flat fl4C M 11 1
l 4
]
l i
! Tigure 2. Plant Response to Generator 1.oad Rejection j Without Bypass (BOC3 to EOC3) l 1 l i
I I, 16 I I
, , 23A5841 REV. 0
. 00 84 ISS. 8 i
! =f N kVI IVIIII. 80fII IIIII8III 2 a,'.g e:
ti $Jr ,
t 44 ft.: 1 3 188 ? vabei 8tt gr' vatig 84:e at:
i d
3 :e( (gj ' FL:y M i i ,l g,3 -q.'; ** 4 BfPill vaLv FL:s l
III.9 .
t il. 8 '
g 'l l [
t 14.5 !
g n ;;
I- ;G .
; ; .- .i::: : :
.. j l '
... 1 ll. 8 28.8 t. 8 18.8 10.3
- t. 8 fler tst :<st f!PC (18C:41)
L ti::: stact: 't
- Ik!*
its t. i(1% =.8tr.
p re o.:.lt*.lasti - ve.t= sti: :v tv -
i s ,, , - 2':!!! $ 1*I' $ ' ,,,,
- i. : } :Ii!!" '!'3!1.!!' (
' j i l
l \
l I
C
.. . . .. 11 ? ,. , .
) / ,
- y :
i 3
- 'I S t. / +""'! I W .i.4 i
fY l
- i k
... ... j L. ...
- 8. 6 10.6 f t.4 f. 8 18.4 36.'
flot altti g st flag agt:<t3 1
I I
Figurs 3 Plant Response to Teodwater Controller Failure, l (50C3 to EOC3) ;
i k
11 l t
I
o .
23A5841 REV. 0 '
00 85 :
i u
9 I ntltt'. parls etst(ptg3
) t =tyte0N tus 1
3 f($9 I
l 2 ait 3.ar actj m( Af PLV8 2.:ncrsetrc:, ,,,,, I,*Saf(M,g!!! .f a.p{uSgy;;
.3,,,
)
3 I 1 ri . .
i i si. -
i 8 'W y' g & '
i 1 I
l ss,e =*
18.8 7 '\ !
\
i
- s. 3 1
i,
,4:-
[ : : :
i 4.8 4.8 8.8 2. 8 4. 8 8.8
- 3. 3 2.8 I !!*1 titC N il ?! M tit:CaCl) {
n 1 4 i
~
. r
' i i.t .t v010 sta:f Ntfv
}
- ellli.
- !%c.str.igt.saaf) 5 *t a m*'
- e I h CC'esta statt!vt77 f L 2 E !'" !!'!!!'t
A a ,, ! 3 ::!!!1}T*":' _
i 3
( -
/ [
A
\ / \
a
. W^
'M 8. 4 ,
181.8 g%
~
t 1 i I t
{
i J
m/t 1. 1 l
f.
Q N l l
- i. . . . .
9 -3 I
.iii. e .,, hi l
- )
t 4.8 3.8 4. 8 4. 0 48 3.8 4.8 4. 8 ;
1 vi.: cu::= cia ein inext:
j i
I 1
l 1
1 l Tigure 4. Plant Response to Generator load pejection Without Bypass (Extended Load Line 7,4 ')
1 18 i
23A5841 REV. 0 ,
00 86 l
l sie.:
C i .tiir. es s erie:essi
- i..ti Sr.> t ... : i.ei n resp ..ci.c et:.
serir: -c . , a w. }sggt[ 6 (f(;,
' # LOW 4
, Ill.I
) (*8l (atl 4 l'ents sage 70s 6
f r lit.8 :
i
$ 188.9 --
- -I.\
- a
- : : -
10.0 g
14.8 ,
I = w- ;; ,
1
- 8. 8 .,_..L.
^
- 4. 8 .
28.1 f. 8 11.4 28.8
- f. 18.8 f!'( t tt PCl) f!*t ist :W ll {
g ,
P l
8 v0!; age ftig*,
tLDttt!%C C EF.t * !
3v(.aslit.l't.**'0,E8.lnat)
- i x i c..-<,:.
t.8 I. ;s.:;.88.ts
- n. - .
.e 1.r n *;;,tt?
at a gt g. : p *2" ,
f,
'\
1 (
I 188.1 li
) g w
- 8. 8 -
.y' -
I I
5 kik- .l.4
\ !
18.3 i
/ Y%
0 t J ... t .:., ,
4, 8 i t. e e. s i t. : Is.:
)- flot tltttscll it.
ttet tsg::<g3 i
1 1
1 I
s 1 Figure 5. Plant Response to Feedwater Controller Failure -
l (Extended Load Line Limit) i i 19 t
,' u Ass 41 arv. 00 87 t
a
.(150. persa eist:Psta d i me.ste: rs.s.< a t eq.g 4 8 5 *'attr 2 test?, va6 4 as:
g;ag :P4,(7 TL e 3 *ti't8 u v44 --
tyv
- ,,, s see.e ;
i
< i i,.. , M> I :....
1
'I*
D w. ..
l
! :. iii.
\
- ... k. ... __
i 4.8 6.8 8.8 2.3 4. 8 8.8 l 3. 8 2.8 !
t!st (St ;tt! ?!*t flit *tt!
f i
i .
i l
I 5
{Littg(N=.8tr.l(P.584f3 l 1 ,01 statt14 fr 3....:k{y....!vt i.
s o 5... m .,(a.
.. v .. i i ...., L i ... .! ! ri !
V
'Q /
h * :
/g1 ' l \N : .
u, %'" h j i. : ) i : I _ w .i.. t
] W l l
i I i
i 188. 8 .g.s ,
f.4 4.8 61 0. 8 3. 0 4. 0 6.8 i 1. 8 v:n istn<s e:n su n <ii i '
4 l
1
\
a ;
4 l Figure 6.. Plant Response to Generator Load Rejection
- i Without Bypass (Increased Core Flow)
- i 20 A _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ ._ __-
MA584 '
00 88 I
L ise..
w.t.> a : <ti .r. ..n FL; t, vag,s ein:esn f
L
' audi $/r 't da twv ilas e i eg ll t. $
Y f- '
is..
Il l
l iii., -
-c M i
i -
g t
14.8 .
g I
w1 . : : l 18.8 _
j
.;. . ; . 3 .: : !
- 8. 9 '
i i 8. 8 29.8 1. 8 18.1 .t.t '
- 8. 2
'!*( t tttlen ttPt 'HC:Til .
t l
i h
ts
- .tnet.
itt t !:'s.c..a a <. :.rt.sts.sn e r)
- ::**.te
.::: sta:
et.vjfv criert, .
is ,, ,
?Ih!!L % !{l' $ ' .,
- H!!" !!'ji!!!!' ,
I E
I -
I / t I- h y t I- .. . .
\/ i
- i
.i.:
s i. :
rw
/ t ,
l 1
1 84 1,8 l
- g. , ;
... i i. . n. . ... n.. .... I ti. t u c:<n ti s tue:<n Figure 7. Plan) Response to Feedwater Controller Failure .
(Increased Core Flow) 21
. O
. 23A5841 REV. 0
- 00 S9 r.. a:st:=s::
st. si i .r. .* ..
i sr.
- . e. ri..,
e.w .ct: -i t s . =.n r.:.
2 ::st t ut F.: ,,,,,
3!t.:H ... ! ' ;<
l I
a lt. -
188.8
! Ok a
I i C
- s 8. 3 l let 6 l-
\ .
I. : 1. 8 _ , .
- 2 .3 4.8 0. 8 8. 3 2. 4 9. 8 8.5
?:=t :st::< t) ' *( !!! :<ll
. .::: as.: :
i .itsa. 5 t. <.:.
.t .t. t : :..s tr.its.; en l :
x :a::ses;!.as it.u c-
!:vtv- !!.: : iv: v 1
^
I >
1 . p 4.! t**<.:.
., l 4
!o.i 4 y 4 3 r. . u_ ! ! 33 -a.
-1
% V b.
N I^
[
[f . -
=m -
t .; I g
l
.618.2 .l.3 3.1 3. 8 4.5 4. 4 1.8 f. 4 4. 8 6.4 i
't c tic::<s r: < u::<ii )
1 i
l l
l Figure S. Plant Response to Generator Load Rejection f
Without Bypass (Increased Core Flov and Final Feedvater Temperature Reduction) 22
.. . . . .- . . . . -. -- . . ._ . . ~ . . . - . . . . - - - . . . _ _ - --
l
- 33A5841 REV. O ;
i I !
j- l ite.s r
.tll g %t rs5.*r* as t'.'s:
I d et rg. t** J. "tll u *!
84I:811) e, a :st ; ,,( t'. :.
3 est at. Er' va vap , ;
" * *
- e g ese gg +.eg . ,
111.3 l ..... _.._
/\%
" 7
-J*
l i
= t e s. '\\ t
. E
- 14 8 i1 ,
. 3
! i
- 10.0 i IV ^
t
... .J ;
I :
I i
- i. _ _ ,
- 5. 8 13.8 19.8 8. 3 18.8 28.3 y
?!*t tit :'Cl3 t!*t tit :tli 4 ,
\ _
i aget*;.; r 4
-L;%c s t str.;ts.;est) e.:.
2..gf'; s. ts ac. ::.:- t it!.!
- 1' '
t.1 !5! -
' ( '
l f
$ t, a 5 .. . , - i / ;
,Rf
. . t
's j\ 5 g f
- . : h.i.
I
/ Q\j [ \
l h
88 1 - _ 3.:
- 3. 3 18.8 28.0 1. 0 t s.1 21.3 f!*t tit :iiCll t;et t tg tCl)
Figure 9. Plant Response to Feedwater Controller Tailure i I
(Increased Core Flow *and Final Feedwater Tenparature Reduction) i
, 23 i I
_ - - - _ - _ _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _l
- . ~_ -. . ._ ..~. . - . - _ . . . - - .
, 23A5841 REV.0 !
00 91 i
$ i
[
2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 ;
59 18 30 30 la 55 14 8 8 8 14 [
51 16 20 0 0 20 18 h 2 ,
! 47 14 8 30 30 30 8 14 '
43 18 20 4 10 10 4 20 la 39 8 30 26 44 26 30 8 -
l 35 30 0 10 0 0 10 0 30 'l 31 8 30 44 16 44 30 8 27 30 0 10 0 0 10 0 30 I 30 26 44 26 30 6 i 23 8
. i li 18 20 4 10 10 4 20, la i
[
15 it 8 30 30 30 8 14 :
I 11 18 20 0 0 20 18 !
l 7 14 8 8 8 14 l 3 18 30 30 18 l
[
f NOTES:
}
- 1. tio. indicates numoer of notches withdrawn out of 48. Blank is !
4 Withdrawn Rod. [
l
- 2. Error Rod is (26, 35). l I
l nwr. w. u=1ein. Ra r ttorn 14 1
- . ._ . .. - . .. . . . .- ~ _ _ . - _ _ _ _ _ .
4 ;
23A5841 REV. 0
)
00 92
- r i
I T i I
i
- (vfem Flys illit. **tly a t g gesg3 ;
I g aug s,statt dat fut l1*tt'v 3:08C l*4.C FL s , J e,(g:gt .vaLC' '. o v tw*, ;
j , _
4 6 1 l 4 i s s. : vd' t es, e 3 % % ,
4 3 4 '3 i I I
- E '
a 18.8 100,8
' s l -
l t
- ... _ ... . ._, . . l j 8. 0 1.0 4.8 1.0 f j f!*t (58:0481 fist tt!!; < tl l
4 F 4
i -
I 6tetLC;*eO=.8tt.5LD.lsera , aTic ngt, ,
eu e g a:t ;
- 2 viitt-. s t* a
- . 1 8!i'! % Fl'M*;' i, I 3 ti!'" f t.s:!.tutte ;p / 1 b
6i a
! ! l
< t is. : = s.: _ .. .
=
r.-
(
\ ,
E i w i j
'+8 N .t. s d I t 1 ...... . . , t !
- l. 8 S.8 4,8 3,3 i '
ttet tltt;icll etag ggg;;ggg
)
l i
a ,
J l
J 1
1 Figure 11. Plant Response to MSIV Closure (Flux Scram) i.
4 I
j :st.:s h
23A5841 REV. 0 00 93 APPENDIX A TRANSIE.VI ANALYSIS BASIS no Loss of Feedwater Heating event was analyzed using the BWR Sinalator Code (Reference A-1). The use of this code is permitted'in l GESTAR II (Reference A 2). I The Loss of Feedwater Heating Event plot normally reported in Section 10 is not an output of the BWR Simulator Codot therefore, this plot is not included in this docume.nt, t RITERENCIS l
I A-1. "Three Dimensional BWR Core Simulator " NEDo 20953A. January 1977.
i A 2. "General Electric Standard Application for Reactor Tuel." !
NEDE-24011-P A-8. May 1986. l l
f i
l l
l 27/28
- 33A5861 REY. 0 l, 1
00 94 APPENDIX 5 l
- GETAB AND TRANSIEVr ANALYSIS INITIAL CONDITIONS i f
The values used in the GETA3 and Transient Anal / sis are given in Table '
B-1. The ic11owing values differ from the values reported in ;
I NEDE-24011-P-A-8-US. May 1986. t i l i Table B 1 !
4 l
2 PLANT PARAMETER l
P Parar.eter Analysis Value NEDE-24011 Value 3323 3453 i Thermal Power MWt 0 14.30 14.97 f
) Rated Steamflow x 10 lb/hr Dene Pressure psig 1006 1020 j
! Non-Fuel Power Fraction 0.038 0.04 L i
- L I Dual Mode' Safety / Relief Valves (18) 1 '
Relief Mode Lou Setpoint psig 1087 1076 I
Safety hode Low Setpoint psig 1162 1165 l i !
1 I
a i I
l 1
I :
l !
1 \
a 1
l l
l I
1 1
l 29/30 1
(Final)
L
--v.-w--, , , , - - e-----,,,-,,-r <-.,-,..-,---wa- ,.,--.-,,--------n,_. _ - -n-m----n ,-,v_------e-.wm- ----,,-m- -,y,w-n--.,ne-m<
00 95 ADDITICNAL INFORMATION REGARDING THE SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR LA SALLE 2 RELOAD 2/ CYCLE 3 Section 3. REFERENCE CORE LOADING PATTERN Cycle N 1 Incremental Exposure 8200 Wd/ST Cycle N Exposure Increment 7450 Wd/ST Cycle N Full Power Capability (if different from above)
Section 4 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTDi WORTH (NO VOIDS. 20'C)
Cycle Incremental Exposure Corrt.sponding to Minimum Shutdown Margin R Value 6000 WD/ST l Section 6. RELOAD-UNIQUE TRANSIEh*T ANALYSIS INP1JT (REDY EVDTTS) ,
EOC Void Fraction (Haling) REDY Not Used EOC Bypass Flow Fraction REDY Not Used Delayed Neutron Fraction (BOC/EOC) REDY Not Used Void Coefficient X 103 (BOC/EOC) REDY Not Used
[ Units are (Ak/h)/(4% voids))
I Section 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS I
PV PV Limiting Power Flow Flux Q/A PSL (Dome) (Botton)
Exposure Transient (% NBR) (% NBR) ( NBR) (% NBR) (psin) (psin) (psin) l EOC TTNBP 100 100 34 2.6 112.8 1141 1149 1177 EOC LRNBP 100 100 Jh 114.7 1143 liso 117m EOC MSIVF 102 100 4 84.2 129.1 1227 12 32 1264 Analy zed MSIVD ,
, (PANACEA) LWH JO 87 119.4 119.2 N/A 1006 1027 est EOC WCF 100 100 238.1 111.1 1122 1128 tiu ,
1 i
e I
00 96 ADDITIONAL INFORMATION REGARDING 11E SUPPLEMENTAL RELOAD LICENSING SUBMITIAL FOR LA SALLE 2 RELOAD 2/ CYCLE 3 Section 10. CORE-WIDE TRANSIENT ANALYSIS REPORT (CONTIhVED)
Were resolved OPL-3 values used for safety Yes and relief valve characteristics?
Assumed MSIV Closure Characteristics:
Time (see) MSIV Area (per unit)
.i 0.0 1.0 (fully opened)
O.6 1.0
- 1.7 0.01 3.0 0.0 (fully clos;d)
A Section 14. ROD BLOCK LINE EQUATION
]
RB 10.58W + 50 l '
.1 Section 15. LOADING ERROR RESULTS l Bundle Type for Limiting Misorientation: N/A l
l
. i a
j
00 9*7 ADDITIONAL INFORMATION REGARDING THE SUPPLEMENIAL RELCAD LICENSING SUBMITTAL FOR LA SALLE 2 RELOAD 2/ CYCLE 3 d
Section 3. REFERENCE CORE LOAD 7NG PATTERN Cycle N-1 Incremental Exposure Cycle N Exposure Increment Cycle N Full Power Capability (if different from above) 1 Section 4. CALCULATED CORE EFTECTIVE MULTIPLICATION AND CoffrROL SYSTD4 WORTH (NO VOIDS. 200C) l Cycle Incremental Exposure Corresponding to Minimus Shutdown Margin R-Value
\
Section 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPtJT (REDY EVENTS)
(Provided unless generic LOTH is reported)
EOC Void Fraction (Haling) i EOC Bypash Flow Fraction Delayed Neutron Praction (BOC/EOC) l Void Coefficient X 10 (BOC/EOC)
[ Units are (Ak/h)/(a% voids))
j j Section 10. CORE-WIDE TRANSIDir ANALYSIS RESULTS/ELLLA .
l l W W Limiting Fower Flow Flux Q/A FSL (Dome) (Botton)
Exposure Transient (% NBR) (t NBR) (% NBR) (% NER) (psin) (psin) (psin)
I j Eoc TTNEP 100 87 290.6 110.8 1142 1150 1174 EOC LRNBP 100 87 327.9 112.7 1142 1150 1174
! EOC MSIVF 102 87 458.7 122.4 1228 1232 1259 b!1ered wstvn a Not j Analvred L7WH i EOC *WCF 100 87 210.6 109.4 1122 1128 1152 p --- -- .-w - - - - - - - - , - - , , _ . , , . - - . - - - -
--c-n- . . - - - . - - - - - . , ,
. 00 38 ADDITIONAL INFORMATION REGARDING THE SUPPLEMENTAL RELOAD LICENSING SVlMITTAL FOR LA SALLE 2 RE14AD 2/ CYCLE 3 Section 3. REFERENCE CORE I.0A')ING PATTERN Cycle N-1 Incremental Exposure Cycle N Exposure Increment Cycle N Full Power Capability (if different from above)
! Section 4 CALCULATED CORE EFFECTIVE WJLTIPLICATION AND CONTROL SYSTEM WORTH (No V0 IDS. 200 0)
, Cycle Incremental Exposure Corresponding to Minimum Shutdown Margin R-Value i
Section 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPtTT (REDY EVENTS)
- (Provided unless generic 14FH is reported)
EOC Void Fraction (Haling)
- EOC Bypass Flow Fraction Delayed Neutron Fraction (loc /EOC)
J Void Coefficient X 103 (30C/?DC)
(Units are (ak/h)/(AZ voids))
l Section 10. CORE WIDE TRANSIENT ANALYSIS RESULTS/ICF :
i !
FV FY l Limiting Power Flow Flux Q/A PSL (Dome) (Botton) '
Laposure Transient LNBj (% NBR) (t NBR) (% NBR) (psin) (psin)
ITN5F 100 105 372.7 113.9 1141 1149 M !
I E0C+190 1178 I E'd /S T g,g vo LRNBP 100 105 413.9 115.3 1142 1150 1180 gyPO MSIVF 102 105 489.1 127.2 1228 1232 1267 N5t Analvsd MSIVD ,
- Elysed LWH _
g go WCF 100 105 259.7 112.3 1122 1128_ __ 1157 I
4
_ _ _ . -. _ - , ~ , _ , - _-- - - - -. . - - - _
00 99 ADDITIONAL INFORMATION REGARDING THE SUPPLEMENTAL RE14AD LICENSING SUBMITTAL FOR LA SALLE 2 RELOAD 2/ CYCLE 3 Section 3. REFERENCE CORE LOADING PATTERN Cycle N-1 Incremental Exposure r
, Cycle N Exposure Increment Cycle N Full Power Capability (if different from above) ,
Section 4. CALCULATED CORE EFTYCTIVE MULTIPLICATION AND CORTROL SYSM WORTH (NO VOIDS. 200C)
Cycle Incremental F.xposure Corresponding to Minimum Shutdown Margin R-Value Section 6. RE14AD UNIQUE TRANSIENT ANALYSIS INPUT (REDY EVENTS 1 (Provided unless generic LOTH is reported)
EOC Void Fraction (Haling)
EOC Bypass Flow Fraction Delayed Neutron Fraction (BOC/EOC)
Void Coefficient X 103 (BOC/EOC) ;
(Units are (Ak/h)/(A% voids)) !
Section 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS/ICF+FFVTR PV PV Limiting Power Flow Flux Q/A PSL (Dome) (Botton) ,
Exposure Transient (% NBP) (% NBR) (% NBR) (% NBR) (psin) (psin) (psin) '
E0C+817 TIWBP 100 105 341.4 112.5 1137 1145 1173 MJd/ST ggyg{L7 LRNBP 100 105 373.6 114.3 1139 1146 1174 gggy7 'MSIVT 102 105 412.0 125.4 1220 1225 1258 N$alyzed MSIVD
$$alvred LFW l ggf7 FWCF 100 W 276.2 116.8 1116 1121 1149 1
00100 REFERENCES 1 1. Unit 1 Cycle 3 Reload Licensing Submittel and Technical Specification Changes, LOSR 87-74 1
- 2. OE Document NEDE-31455, ' Extended Operating Domain and Equipment Out of Service for LaSalle County Nuclear Station Units 1 and 2', dated ,
November 1987 with addendum in Attachment 6. t l
- 3. GE document NEDC-31510P, 'LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Less-of-Coolant Accident Analyses,' dated December 1987, Proprietary.
i f
I l 1
t i
u I
4 l
l r
a l
I l
{
I !
l I
- 1 L I e .
i
V~ ! L) I ATIACIMEffT_G GE Document NEDC-31510P, "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analyses", Addendum Issued March, 1988, Proprietary.
This document will be sent under separate cover 5102K 1
1 i
l
e s t ') $
ATIACIMENT_H GE STANDARD LICENSING STABILIW ANALYSIS RESULTS i
1 1
I l
5102K i l
1 l
l 1
J t
l ,
.l.
.- 00103 i
~
GE Nuclear Energy .
( l W:,J,,': .,' . , , . . . . . \
i August 9, 1988 cc: W'.F. Naughton REP:88 173 i Mr. R. A. Roehl i Supervising fuel Buyer 1 COMMONWEALTH EDISON COMPANY Fuel Department, 234 E P. O. Box 767 !
l Chicago, IL 60690 l (
1 l
SUBJECT:
LaSalle Unit 2 Reload 2 Stability Analysis !
REFERENCE:
- 1. "Contract between Commonwealth Edison Company and General Electric Company for Fuel Bundle Fabrication and Related Services for Quad .
Cities Nuclear Power Station and LaSalle i County Station," dated January 6, 1986, i
ATTACHMENT: LaSalle Unit 2 Reload 2 Standard Licensing Stability Analysis i
Dear Mr. Roehl:
Attached are the results of the standard licensing stability analysis for LaSalle Unit 2. Cycle 3. These results are based on approved GESTAR Il models I using approved procedures and were calculated in the same way as the Cycle 2 results. However, if LaSalle has a pump trip event similar to that expert-enced in Cycle 2, they should expect results similar to those observed in ,
Cycle 2. :
Very truly yours, P
R. E. Parr Senior Fuel Project Manager l Edison Projects i l M/C 174; (408) 925 6525 j REP:mg.3 S
1.
I
, .f l . i
, , 001"4 ,
i I
ATTACHMENT 1 '
)
LaSalle 2/ Cycle 3 l The results of the standard licensing stability analysis are given below,
- consistent with the cycle 2 analysis approach. The results reflect the t
- effects of the more negative C 3 void coefficients and the lower pressure !
drop GE 8 fuel, i
! DECAY RATIO AT THE INTERSECTION OF NATURAL CIRCULATION AND 105% Rod line Extran. APRM Rod Block
]
Core Decay Ratio .63 .72 ;
I l Channel Decay Ratio j P8x8R/8X8R .47 .63 '
- GE 8 .42 .56 l
- l 1
j The plot for this case is shown in Figure 1. ,
F j
f r
i I
I I
t
1 l
. . ., i 00105 I
. . i l
A NATURALI CIRCULAT! JN 9 105 PERCENT RUD L INE C ULT. PERFORMANCE _IMIT 1.00 C :
.75 s A N
- x C
" .50
>- \
u \ '
.'s.
, 1
. .v
\
. . .~u g -
- 0. 0 20.0 40.0 60.0 80.0 100.0 12: 0 :
l PERCENT POWER l
Figure 1. Reactor Core Cecay Ratio vs Power l
l
% __ _n_