ML20148N411
| ML20148N411 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 01/31/1988 |
| From: | Charnley J, Elliott P, Lambert P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20148N405 | List: |
| References | |
| 23A5846, 23A5846-R, 23A5846-R00, NUDOCS 8804060506 | |
| Download: ML20148N411 (20) | |
Text
_ _ _ _ _ _ _ _ _
O 23A584G REVISION 0 CLASS I JANUARY 1988 0
23A5846 (REV.0)
SUPFLEMENTAL RELCAD LICENSING SUBMITTAL O
FOR QUAD CITIES NUCLEAR POWER STATION UNIT 2, RELCAD 9, CYCLE 10 g
i i
Prepared by:
J P. A. Laddert Fuel Licensing Q
(
i f0 Verified by:
P. E. Elliott Fuel Licensing O
/
Approved,71
\\
J,
. Charnley p'
F neger, Fuel Lice..ng O
O 8804060506SSoM65 1/2 PDR ADOCK PDil P
O 23A5846 REV. 0 l
IMPORTANT NOTICE REGARDING
- ()
CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Commonwealth Edison Company (Edison) for Edison's use with the U.0. Nuclear Regulatory F
' C)
Commission (USKRC) for amending Edison's operating license of the Quad Cities Nuclear power Station Unit 2.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
()
The only undertakings of the General Electric Company respecting information in this document are contained in the contract between Commonwealth Edison Company and General Electric Company for fuel bundle fabrication and services for Quad Cities Nuclear powee Station Units 1 and 2, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract,
- {)
or for any purpose other than that for which it is intended, is net authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as tv the completeness, accurary or usefulness of the information contianed in this document or l
that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind
. C) which may result from surh use of such informatten.
2()
4 1
C)
O O
()
O 23A5846 REV. O O
A xsowntoarxts s The engineering and reload licensing analysis, which form the technical basts of this Supplemental Peload Licensing Submittal, O
were perforces in tne.vuelear ruel and Engineering services Department by F.. J. pearsen.
O O
O 3
O O
O O
O s/6
23A5846 REV. 0 1.
PLANT-UNIQUE ITEMS (1.0)*
Analysis Conditions:
Appendix A 2.
RELOAD TUEL BUNOLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Type Cycle Loaded Numbe r Irradiated
)
TBDGB263L **
6 8
P8DGB298 **
6 24 BP8DRB265H 7
200 BP8DRB282 8
72 DP8DRB283H 8
lu4
)
BP82RB299 9
64 BP8tRB299L 9
BB New BD3000 10 92
)
B0316A 10 72 Total E
)
3.
REFERENCE CORE LCAOING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycles 21666 mwd /MT Minirum previous cycle core average exposure
)
at end of cycle from cold shutdown censiderations: 21335 MW3/MT Assured relcad cycle core average expcsure at end of cycles 22754 M'3/MT Cere Leading pattern Tigure 1
)
a( ) Refers to area cf in discussien General Electric Standard Application for Reacter Tuel NE0E-240ll-P-A-8 (dated May 1966): a letter "S"
)
preceding the number refers to the United States Supplement.
Barrier fuel.
==
7
)
l
[
O 23A5846 REV. O O
4.
CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM VORTH-NO VOIOS, 20 CE3. C (3.3.2.1.1 AND 3.3.2.1.2)
Beginning of Cycle, L-effective O
Uncontrolled 1.102 Fully Controlled 0.956 Strongest Control Rod out 0.961 O
R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, Delta k O.007 6
5.
STANOBY LIQUID CONTROL SYSTEM SHUTDCVN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Delta x)
O EE3 (20 deg.C. Xenen Free) 600 0.043 9
O O
O B
O
O 23A5846 REV. 0
()
6.
RELOAD-UNIQUE TRANS!ENT ANALYSIS INPU* (3.3.2.1 5 AND s.2.2)
(COLD WATER INJECTION EVENTS ONLY)
()
Void Traction (%)
34.19 Average Tvel Temperature (degrees F) 987 Void Coef ficient
()
N/Ae (cents /4 Rg)
-4.61/~5.76 Doppler coefficient N/A* (cents /deg. T)
-0.244/-0.232 Scram Worth N/A* ($)
O Extended EOC with Increased Core T10w and Final Teodwater Temperature Reduction
()
Vo;d Traction (4) 28.50 Average Tuel Temperature (degrees F) 983 vota coefficient
()
N/A* (cents /n Rg)
-5.04/-6.29 Doppler Coefficient N/A* (cents /deg. T)
-0.273/-0.259 Scram Worth N/A* ($)
O O
- N = Nuclear Input Datas A = Used in Transient Analysis
- Generic @xposure-independent valuts are used in Gener44 Electric Standard Application for Reactor Fuel, NEEE-24011-p-A-0. May 1996.
O 9
0
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23A5846 REV. 0 9
7 RELCAO-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Fuel Paaking Factors R-Bundle Bundle Flow Initial Design Local Radial Axial Factor Pewer (MVt)
(1000 lb/hr)
MOPR O
Exposures B0010 to EOC10 BP/F8xBR 1.20 1.83 1.40 1.051 d.206 104.0 1.28 GE8x8EB 1.20 1.C*
1.40 1.051 224 106.7 1.28 O
Exposure: Extended EOC with Increased Core Flow and Final Feedwater Teeperature Reduction BP/P8x8R 1.20 1.82 1.40 1.051 6.159 114.7 1.3A GE8x8EB 1.20 1.83 1.40 1.051 6.177 116.7 1.31 0
8.
SELECTED MAR 3IN IMEROVEMENT OPTIONS (S.2.2.2) 9 Transient Recategorization:
No Recirculation Puep Trips No Rod Withdrawal Limiter:
No Thermal Power Monitor:
No Exposure Dependent !.imits:
No g
Exposure Points Analyzed:
1 9.
CPERATIN3 FLEXIBILITY OPTIONS (S.2.2.3) g Single-l.c0p Operation:
Yes L0ad Line Limit Yes Extended Lead Line Limits Yes Increased 00re Ficw Yea Flow Point Ana;yzed:
100%
II Feedwater Temperature Reduction:
Yes ARTS Progras:
No Maximum Extended Cperating Domain:
No 9
10 O
i O
9 23A5846 REV. 0 10.
CORE-WICE TRANS!ENT ANALYSIS RESVLTS (S.2.2.1) t Methods Used: OEMINI f
Taux
-Q/A Delta CPR I
(% M8R)
(% NBR)
SP/P8x8R GE8x8EB Tigure i
O i
Exposure SOC 10to E0C10 Turbine Trip w/o Bypass 535 122 0 '.
0.21 2
[
Inadvertent HPCI 121 116 0.18 0.18 3
O Taedwater controller Ta11ure 280 118 0.15 0.16 4
l 1
I Exposure: Extended Eoc with Increased Core Tiow and Tinal Tes$ water Temperature Reduction Load Rejection w/o Bypass 475 121 0.24 0.24 5
i Inadvertent HPO!
118 114 0.14 0.14 6
[
4
(
Teodwater Controller Ta11ure 276 122 0.18 0.18 7
l 40 i
l t
2 l
l 11.
LOCAL ROD WITHDRAWAL ERROR (WITH L14! TING INSTRUNENT TAILURE)
(Q TRANSIENT
SUMMARY
(S.2.2.1)
{
)
Limiting Rod Pattern Tagure 8 i
i Rod 3*ock Rod Position telth NLHGR O
('*** **'**")
'""*'Y'*"-
BP/Paxen arexers j
}
' 104 3.0 C.07 15.59 16.59
[
105 3.5 0.10 16.61 17.6!
4 106 4.0 0.12 17.73 10.23 107 4.0 0.12 17.23 10.23 f
i 108 7.0 0.20 17.26 10.26
(
- O 10, 12.0 0.26 17.26 18.26 110 12.0 0.26 17.26
..B. 2 6
[
]
[
]
i Setpoint selected:
100 0
11
[
O t
i
O 23A5846 REV. O GD
- 12. CYCLE MCPR VALUES (S.2.2)
Non-Pressurization Events Exposure Range: BOC10 to EOC10 BP/P8x8R GE8x8EB Inadvertent HPCI 1.22 1.22 Fuel Loading Error 1.18 9
Rod Withdrawal Error 1.24 1.24 Exposure Range Extended EOC with Increased Core Flow and Final Feedwater Temperature Reduction II '
Inadvertent HPCI 1.18 1.18 Fuel Loading Error 1.15 Rod Withdrawal Error 1.16 g);
Pressurization Events Exposure Ranges BOC10 to EOC10 g,
Option A Option B BP/P8x8R GE8x0EB BP/P8x8R GE8x8EB Turbine Trip w/o Bypass 1.31 1.31 1.26 1.27/
Q+
Feedwater Controller Failure 1.25 1.26 1.20 1.20 Exposure Range: Extended EOC with Increased Core Flow and Final Feedwater Temperature Reduction g
Option A Option B BP/P8x8R GE8x8EB BP[P8x8R GE8x8EB Load Rejection w/o Bypass 1.34 1.35 1.29 1.30 GD '
Feedwater Controller Failure 1.28 1.29 1.23 1.24 12 GD 4 e
,-er-----
n a-
-.--e,.
n, - =-
m
-e-,- - -
i D
23M846 REV. O j
13.
OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3)
Pressure Pressure Steam Line Valve Plant Transient (psig)
(psig)
Response
MSIV Closure 1300 1324 Figure 9 (Flux Scram)
O 14.
LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes
!Q Event Delta CPR Rotated Bundle Error O.14 15.
C NTR L R DR P ANALYSIS RESULTS (S.2.5.1) 0 Control Rod Drop Accident Analysis is not required for banked position withdrawal sequence plants. NRC approval is documented in NEDE-240ll-P-A-8-US, May 1986.
O 16.
STABILITY ANALYSIS RESULTS (S.2.4)
BWR 2/3 plants are exempt from performing cycle-specific stability
'O analyses.
17.
LOSS-OF-COOLANT ACCIDENT RESULTS O
LOCA Method Used: SAFER /GESTR-LOCA See Quad Cities Nuclear Power Station Units 1 f. 2 SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis, NEDC-31345P, June 1987 (as amended).
O 13 0
O 23A5846 Rev. 0 mMMMMMm eMMMMMMMMM.
eM M ME+8M M M M M M M.
ME+EMMMMMMMMMBEM mMBEMMMMMMMMMMMm
- MMBEMMMMMMMMMMMM
- MEMBEMBEBEMBEMMMBEMM
- ':MMMMMMMMMMMMMMM
'::MMMMBEMMMMMME+8MMM
'::MMMMMMMMMMMMBEMM
" M M BE M M M E+8 M M M M M M
- MEMBEMMMMMMMMM l'
"MMBEE+EMMMBEMMM" "MMMMMMMMM" MMMMM" l IIIIIiliI 1 3 5 7 911131517192123252729313335373941434547495153555759 FLEL TYPE g
A = BP8LRB282 F = BD316A B = BP8DRb283H G = P8DGB263L C = BF8DRB299L H = P8DGB298 D = BP6DR3299 I - BP8DRB265H E = BD300C figure 1.
Reference Core Loading Pattern 14
O 23A5846 Rev. 0 O
l ifEUIRCN FLUX 1 VE5 ;EL PRESS RISt(PSI) 2 AVE SLFACE DE AT FLUX 2 SAF :TY VALVE FL0s
~
3 COR : IP4.[T FLOW 3 REL EF WA,LVE FL0s 159.9 30s,:
i e-m iss u _ur ri r-O
' /(
188.8 O -
0 210.0 g
1 b
Y O
y r
SI. 9 g g g, g N
M O
.._. l _.-
_ 9
- e. '
- 8. 8 19.0 20.0
- 4. 0 10.0 23,3 TIME ISEC0@S) flME ISEC)@SI O
I LEV iL(INCM. RET.SEP.SFRT)
I VO! ) CE ACT!WITY 2VESEEL STEAFFLCW 2 DOF >LER REACTIV!1Y l
3 SC# lM REACilV!1Y 3 TUR ilPC STE,AWLOW 1.8 1
4 vm 6.
eri-etutiv 200.0 e err uiwr r. m.
O 3
g... a
- 1....
1 t
O E
! 6._
E u
v
.l..
W
') '
\\
l 10 168.8 7, g
- 0. 0 13.0 20.0
- 0. 0 19.0 20.0 flME IStCo@s) f!RI Estt0C51 lO Figure 2.
Plant Response to Turbine Trip liithout Bypass I
iS O
O 23A3846 Rev. O O
1NEurRON Flux 1 VESSEL PRESS RISE (PSI) 2 AVE SURF ACE K AT FLUX 2 REL I U VALVE FLCW 3 COR IPLET FLOW 3 BYPASS WALVE FLCW
'm r tu et e:m 150.9
'M r rLy r
- Or r-i 158.8 O
a LA-
-s 1,
W 188.s I'
'/
180.0 g
b
~
5 0
x?
50.4 50.0
.i4 0
i O
88 k
2 2
2 ;
88 M...
- 8. 0 30.8 100. 0
- s. 8 50.0 100. e TIME (SECON0$1 flME (SEco @ SI O
1 LEY:L ( INCw n EF-SEP-SF Ril I volb REACilvliY i
2 VES iEL STEAMFLCv 2 DOP>LER REAtilv!TY 3.0 3 sc,s LM REACT!vliY 31U4 stNE STEAMFLov
'e t u_
nemattutv-150.0
' rer :-avre et y O
?
100.0 AN 5 88
~
l u
0 l
>=
1
?.
)
58.4
/
W
-1.0 Tf r
Y G
- 9 2.8
- t. I 5 B. e 109. O
- e. I 58.8 108.9 ilt (SECO@ll flME (MCO@SI O
Figure 3.
Plant Response to Inadvertent Activation of HPCI 16 9
O 23A5846 Rev. 0 9
sse.o
- e. cura:n vts,tL press nist(Ps!)
2 AVE SLF E
AT FLUX 2 SAF TT VALVE FLCv I-FL v 3 REL EF VALVE Flow 3 COR ? It Is s. O
' e ^= ' ' u_ pE 4 STP L$s VALVE FLOW 180.0 D
'0
-> m,i a
g 10s.0
-=
a b
ss.0 1
/^I E
55.0 0.0
,,. L,-
I. S L
O t.9 20.0 48.9
- 0. 9
- 28. d 40.8 T!nf (EECCS) flME IEttCS)
-O I LEY:L(INN.RU.SEP SMRT)
! v31 ) RE AC 'Ivlty 2 YES!EL STEArrL0v 2 DOP>LER RLaETIv!TY 3 TUS 114 STf *MFL Cv 3 SCG iM REAt Tivity 15 8. C
' r e r
e e er N I*8
' t a' er iet*te e
j
- j O
E q
x -, --
4 l...
.w,-
j.
1
- O
=
41 S t. e il V
-1.0 i
W V
I%
- \\,
g
- e. c 2.0
. e. e
- e. o r e.
4e 8 fimE (SICO G ;
flPC (SEE0@$1 O
Figure 4.
Plant Response to Icedvater Controller Failure 17 O
O 23A5846 Rev. O I Ntut#th FLUX l VE5LEL PRESS RiSt(PSI) 2 Art
$@f A.C1 DE AT FLUE 2 SAF :TT TALYE FLov O
ESe. '
3...
2
^
s.....
e 3
e b
E w2 5..
33..
\\
e M
L
. _.]_
- 8..
l e..
2.. e
- e..
i s..
2...
TisE (ECDet)
Tint (Ecoes)
O I Liv :L(INCw egt.tte.snRT)
- vo}h stattivitt 2 vi$itL STEAmrtov 2 ritstLta etsCliv!!Y
! M !%5'8 'S"'
I
? Et" 5_5!!!!!!
,,s..
s e
E s e e..
le.. y/
E k?
I
,p g...
.\\
v.i..
- e..
t I
G-
.i...
i s..
i..
re.e T!fE (SECDel) fint 19tCCCSI 4
Figure 5. Plant Response to Generator Load Rejection Without Bypass (Extended EOC With Increased Core Flow and Feedwater Te:nperature Reduction) 18 e
I r
23A5846 R2v. O I NE L'rRcn Flux 5 visEEL PRESS RISttPSI) 2 Avt SWACE DEAT Flux 2 Rit lEF W ALnt Fiev Q
3 C O' IPLET FLov 3 DrF iS5 VALvt FLCW 354.9
' * ^ ^
L
- =
33e.e
- P "^'"'
t
' ~
g g g, g g,,,,
O a
i" 39.8 54.9 0
f
- e. t e.g,
r_
- 8. 0 S t. 6 188.0
- 0. 8 50.9 164.9 TIE (ECDell flE IKtosell O
8LtdLtitok2Cr.0r'.0CT) i V01 7 f?CAttiv!TV 2 vil :tL STEanrLCv 2 DC8 "LER PEA 011v!ff-3 tut ilta sita4 Low l
3 sce m Ers:ttytty 158.8
"' ? ?
3*8 E *P"""
O I..
7 %
l..
hs::
=
c: : ::
=
c N
'O E
.t S t. t I.3.8 W
l'O
- 8. 8
- 2. 0
- g. O S t. 8 105.9
- 8. 8 58.0 168 8 j
TIM tl(C M 6) fint (SEC M il l
l t O l
l O
Figure 6.
Plant Response to Inadvertent Activation of HPCI (Extended EOC with Increased Core Flow and Feedwater Te=perature Reduction) 19 O
O 23A5846 Rev. 0 150.0 1 hfuTRON F Ju i vtS!EL PAtti Alst(PSI) 2 AVE 5pr C M AT LUI 2 f a.F E f f VALV[ FLQW O
3 C0AE I
- Lov 3 PELIEF V ALv: FLov E?
4 STF ASS VALVE F60s 35 1.0 W
h h 18 8. 0,
k t
St. e g
E N
- 30.0 W
I
- e..
t l
L
- 9. 0 it.9 28.8 30.0 0.0 19.0 28.8 34.0 T!ME tStCDect3 11Rt tSECDect3 3 LtytLtICLTr.0tr.0K::13 3 V0 D r*Atti'/ ! T T 2 VE5SEL STE AFLOW 2 00PPLIP RfCTIVITY
! E!"' Q:O'!"
' M!"$5'PZ'o' su.
i 4
! i....
k
=7 e.:.. __.
r-f
+
a b
i!
w a
r. s e..
=d
=
v.i..
[
f i
i 1
4 e
s*
l
- e. e i..
2...
ii..
i e..
2..
n.:
714 estcog-53 Itr4 tsttcets:
e Figure 7.
Plant Response to Feedwater Controller Failure (Extended EOC O
with Increased Core Flow and Feedwater Temperature Reduction) 20 0
O 23A5846 Rev. O O
2 6
10 14 18 22 26 30 34 38 42 46 50 54 58 O
59 12 12 12 55 0
0 0
0 51 16 12 16 47 0
10 0
0 10 0
43 16 40 40 16 40 40 16 39 0
10 10 10 10 10 10 0
35 40 44 36 40 36 44 40 31 10 10 0
10 10 0
10 10 0
27 40 44 36 40 36 44 40 23 0
10 10 10 10 10 10 0
19 16 40 40 16 40 40 16 15 0
10 0
0 10 0
0 11 16 12 16 7
0 0
0 0
3 12 12 12 O
NOTES:
1.
ho. indicates number of notches withdrawn out of 48. Blank is a withdrawn rod.
2.
Error rod is (18, 31).
O O
Figure 8.
Limiting Rod Pattern 21 O
0-23A5846 Rev. 0 O
1 NEUIRON FLUX l VE5iEL PRESS RISE (PSI) 2 AVE SURF ACE HE AT FLUX 2 SAF iTT VALVE FLOW 3 COR IM.ET FLOW 3 REL (EF WALVE FLOW 158.9 300.5
' = * *'ee u*
'r et ""
I e
118.0 200.8 5
8 l
I O
r 58.s I.
i so.
~
O
- s. e
- e. e m
- 8. 0 S.8 10.0
- f. B S.0 10.8 TIME (SECCNOS) flME (SECOM)$1 9
i lev :L ( INCH.R E F -S EP.SM R T )
/
I v01 ) REAcilv11Y 2 VES LEL STEAMFLOW l
2 00FPLER REACTIv!TV 3 TUR31NE STE AMFLOW 3*'
f 3 SCE lM REACTIv!TY 200.0
' ' E E '" ' ' E " "! 0"
'~C' '_"E""""
i l
l 9.
3 b ,:
M le s. o 5
i t
M 8
1
- N
/
9-
~
.L
- e. o V.i.e g2 W
T O
. i n.1
-2. s
- 6. 0 3.0 10.9 0.0 5.0 30.8 TIME (SEC0435)
IIME (5(CoeOS)
O Figure 9.
Plant Response to MSIV Closure, Flux Scra:n 22 0
)
23A5846 REV. O APPEND 1X A ANALYSIS CONDITIONS To accurately reflect actual plant parameters, the values listed in Table A-1 were used instead of the values reported in NEDE-24011-P-A-US, May 1986.
)
TAELE A-1 PLANT 7ARAMETERS
)
Parameter Analysis Value NEDE-24011 Value Non-fuel Power Fraction 0.038 0.035
)
Relief Valve capacity, lb/hr 558,000 645,000 Safety / Relief Valve,
)
(
Lowest Setpoint, psig 1135 1135 + 1%
)
2 3
J 23/24 (FINAL)
D
O e
NEDC-31449 Class II July 1987 3
EXTENDED OPERATING DOMAIN A';D EQUIPMENT OUT-OF-SERVICE O
FOR QUAD CITIES NUCLEAR PO'a'ER STATION UNITS 1 A';D 2 H.X. HOANG O
O O
Approved:
App oveC j l'
WF
/
O c.L. S02:1, Ma ;
.r
- Charnley, Application Engineering fuels Licensing Mar.er Services
- O GENER AL @ ELECTRIC 990ga4y 68.
73pp.
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT gu O
Please Read Carefully The only undertakings of the General Electric Cortpany respecting information in this document are contained in the contract between the custocer and the General Electric Company, as identified in the purchase order for this report, and nothing contained in this document shall be construed as changing the contract. The used of this information by anyone other than the customer or for any purpose other than that for which it was intended, is not authorized; and with respect to any unauthorized use, che General Electric makes no representation or warranty, and assures no liability as to the completeness,
- accuracy, or usefulness of the information contained in this docu: tent.
s O
N ii 10
J
!O j
i C)
I TABLE OF CONTENTS
!C) s Pare SDiMARY xi io PART A: EXTENDED OPERATING DOMAIN ANALYSIS FOR QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 A.1 INTRODUCTION A.1-1 A.2 PLAST OPERATIONAL STATUS A.2-1 g)
A.2.1 Recirculation System Capability A.2-1 A.2.2 Minicum End-of Cycle Coastdown Power Level A.2-1 A3 SAFETY ANALYSIS A.3-1 O
A.3.1 Abnorcal Operational Transients A.3-1 A.3.1.1 Limiting Transients A.3-1 A.3.1.2 overpressurization Analysis A.3-1 A.3.1.3 Rod Withdrawal Error A.3-2 A.3.2 Fuel Loading Error A.3-2 A.3.3 Rod Drop Accident A.3 2
')
A.3.4 Loss-of-Coolant Accident A.3-3 A.3.5 Thermal-Hydraulic Stability A.3-3 A.4 MECHANICAL EVALUATION OF REACTOR INTERNALS A.4-1 AND FUEL ASSEMBLY n'#
A.4.1 Loads Evaluation A.4-1 A.4.2 Loads Impact A.4-2 A.4.2.1 Reactor Internals A.4-2 A.4.2.2 Fuel Assemblies A.4-2 A.5 FLOW INDUCED VIERATION A.5-1 r3 v
A.6 FEEDWATER N0ZZLE FATIGUE ANA'
-'IS A.6-1 A.6.1 Methods and Assumptions A.6-1 A.6.2 Results A.6-3
(
A.7 CONTAINMENT ANALYSIS A.7-1 iii O
t NEDC-31449 O
CONTENTS (Continued) faLt PART B: PLANT EQUIPMENT OUT-OF-SERVICE FOR QUAD CITIES
()
NUCLEAR PO'.'ER STATION UNITS 1 AND 2 B.1 INTRODUCTION B.1 1 B.2 FEED'.'ATER HEATER OUT OF-SERVICE B.2 1
)
B.2.1 Abnormal Transients Evaluation B.2-1 B.2.2 LOCA Analysis B.2 3 B.2.3 Feedwater Nozzle Fatigue Analysis B.2-4 B.2.4 Containment Loads Evaluation B.2 5
- O B.3 ONE SAFETY / RELIEF VALVE OUT OF-SERVICE B.3 1 B.3.1 Abnormal Transients Evaluation B.3 1 B.3.2 LOCA Analysis B.3 2 l()
B.4 ONE RECIRCULATION PUMP OUT-OF-SERVICE B.4 1 B.4.1 Minimur. CPR Fuel Cladding Integrity B.4 1 Safety Limit B.4.2 Minimum CPR Operating Limit B.4 2
)
B.4.3 Stability Analysis B.4-2 B.4.4 LOCA Analysis B.4 3 REFERENCES R-1 ACKNOLEDGEMENTS R-2 1
lC) lC)
I
! C) iv (C) l i
$)
.O TABLES Table Ijtle Pare C)
A.3-1 Core-Wide Transient Analysis Results at ICF A.3-5 and ICF/FTWTR A.3-2 Core-Vide ACPR Results at ICF and ICF/FFWTR A.3-6 A.3-3 MCPR Operating Limits at ICF and ICF/FFTR A.3-7 C) for QCNPS Unit 1, EOC10 A.3-4 Overpressurization Analysis Results A.3-7 A.6 1 Feedwater Duty Map Indices Added for FFWTR A.6-4 and Coastdown 0
A.6-2 No::le Fatigue Usage for an 11 Year Seal A.6 5 Refurbishment Period for FF'a'TR and Coastdown (Location I)
B.2-1 Transient Analysis Results for QCNPS at B.2-6
()
100P/100F, Feedwater Heater Out-of Service B.2-2 Operating MCPR Results for QCNPS at B.2-7 100P/100F, Feedwater Heater Out Of Se';vice B.2 3 Feedwater Duty Map Indices Added for B.2-8
()
Feedwater Heater Out of Service B.2-4 No::le Fatigue Usage for an 11-Year Seal B.2-9 Refurbishment Period for Feedwater Heater Out of-Service (Location I)
()
B.3-1 S/RV Setpoints Used for Transient Analysis B.3 4 C
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'O NEDC-31449 O
ILLUSTRATIONS Ficure Title Pare O
A.1-1 Extetid&il Operating Domain Power / Flow A.1-2 Map for QCNPS A.3-1 Plant Response to Load Rejection without A.3-8 Bypass, 1004 Power 108% Flow, Rated Feedwater Temperature, EOCA128 L'D/MT A.3-2 Plant Response to Load Rejection without A.3-9 Bypass, 1001 Power 108% Flow, FFWTR, EOC+582 KJD/MT A.3-3 Plant Response to Turbine Trip without A.3-10 0
Bypass, 1001 Power 100% Flow. Rated Feedwater Temperature, EOC A.3-4 Plant Response to Turbine Trip without A.3-11 Bypass, 100% Power 108% Flow, Rated Feedwater Te=perature, EOC+128 L'D/MT O
A.3 5 Plant Response to Turbine Trip without A.3 12 Bypass, 1001 Power 108% Flow, FFWTR, E0C+582 L'D/MT A.3 6 Plant Response to Feedwater Controller A.3-13 O
Failure, 1004 Power 108% Flow, Rated Feedwater Tenperature, EOC+128 MWD /MT A.3 "
Plant Response to Feedwater Controller A.3-14 Failure, 1004 Power 108% Flow, FF~iTR EOC+582 L'D/MT O
A.3 8 Plant Response to MSIV Closure Flux Scram A.3 15 100% Power 108% Flow, Rated Feedwater Temperature, EOC+128 L'D/MT A.6 1 Example of Linear Method of Determining A.6 6 O
Seal Refurbishment Intervals B.2-1 Plant Response to Load Rejection Without B.2-10 Bypass, Feedwater Heater In Service (Bounding Power Shape)
O B.2 2 Plant Response to Load Rejection Without B.2-11 Bypass. Feedwater Heater Out of Service (Bounding Power Shape) vi O
I NEDC-31449 (3-B.2-3 Plant Response to Feedwater Controller B.2 12 Failure, Feedwater Heater In Service (Bounding Power Shape)
B.2-4 Plant Response to Feedwater Controller B.2-13 Failure, Feedwater Heater Out of-Service (3
(Bounding Power Shape)
B.3 1 Plant Response to Load Rejection Without B.3-5 Bypass. Full Relief Capacity B.3-2 Plant Response to Load Rejection Without B.3-6
()
Bypass, One S/RV Out-of-Service B.3 3 Plant Response to MSIV Flux Scrat.
B.3 7 One S/RV Out-of-Service O
O C)
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O NF.DC.31449
SUMMARY
O This two-part report documents a comprehensive set of analyses performed for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2 to:
(1) Support the expansion of the operating domain of the current O
power / flow map.
(2) Justify continuous operation with one of the following equipment ut-f-servicc:
n feedwater heater string, one O
safety / relief valve, one recirculation loop.
The first part of the report addresses the technical bases to justify the proposed operating domain expansion, and the second part presents O
analyses results and associated technical specification limits to support plant operation with certain equipment out-of-service.
The standard operating domain for QCNPS Units 1 and 2 was previously O
modified to include the operating region above the rated rod line bounded by the 108% average power range monitor ( APPS) rod block line, the rated power line and the rated core flow line (Reference 1).
For this analysis, th'
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flow (ICF) region (Figure A.1 1).
This safety evaluation is also applicable to operation beyond nominal end of cycle with ICF and up to 100 degree F final feedwater temperature reduction ( FWTR) or beyond nominal end-f- ycle at less than rated c re f1 v vith reduced feedwater O
temperature.
The cycle extension is then followed by a coastdown to 20i power.
As part of the expanded operating domain analysis, the limiting O
abnormal operational transients at rated core flow condition are core flow with and without FWTR to reevaluated for 100% power and 108%
the ICF operation mode throughout the cycle and ICF/FWTR operation support O
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beyond nominal end-of-cycle.
The operating limits obtained for ICF with FF'm7R also bound operation at less than rated core flow with reduced O
feedwater temperature.
The loss-of-coolant accident (LOCA), the containment LOCA response,
- O and the reactor stabi)ity compliance criteria were also evaluated to justify operation at ICF/FF'7R conditions.
In addition, the effect of the increased pressure differences due to ICF on the reactor internal components, fuel channels, and fuel bundles was also analyzed to show that their design limits will not be exceeded.
The effect of ICF on the flow-O induced vibration response of the reactor internals was also evaluated to ensure that the response is within acecptable limits.
The increase in the feedwater nozzle usage factors due to reduced feedwater temperature was
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Results of the analysis reported herein show that there is no impact on LOCA performance, c catainment design loads, reactor internal loadings
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expanded po'-er/ flow domain.
The recirculation system performance data for QCSPS indicate that the currently achievable core flow rate may be less than 10Si ef rated.
However, this is a system performance consideration rather than a safety concern.
Therefore, the analyses were performed at O
the bounding condition of 108% rated core flow.
For the out-of-service equipment mentioned in Item 2 above, the analyses performed assumed a single failure only and established the O
licensing bases for continuous plant operation in the expanded power / flow map excludinn the ICF rerion.
'a h e feedwater heater out-of-service (corresponding to a 100 degree F reduction in feedwater temperature) is included as part f the transient analysis input assumptions.
spec.fic O
cycle independent operating MCPR limits are established to allow continuous plant operation with this equipment failure.
In the case of a
safety / relief valve (S/RV) out of service, transient analysis results O
X O
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NEDC-31449 showed that there is no impact on the calculated MCPR and that the overpressura margin to the ASME upset code limit is still satisfied.
An analysis justifying recirculation system single loop-operation (SLO) was previously performed for QCNPS (Reference 2).
This analysis has been reviewed and shown to be applicable with the new GE8x8EB fuel design.
In O
addition, the impact on plant operating limits were also evaluated for SLO mode with or without one safety / relief valve out-of-service in the normal operating domain as well as in the region above the rated rod line.
The analyses of the above-mentioned equipment out-of service also showed that
=O there is no impact on the LOCA containmer*- response, reactor stability performance, or fuel peak clad temperature.
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1 PART A i
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EXPANDED OPERATING DOMAIN ANALYSIS t
FOR QUAD CITIES NUCLEAR P0k'ER STATION UNITS 1 AND 2 l
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.O NEDC-31449 A.1 INTRODUCTION O
Most boiling water reactors (WRs) have recirculation system purnping O
capabilities in excess of that required to provide 100% rated core flow.
The use of increased core flow (ICF) above 1004 rated core flow can provide greater operational flexibility in reaching and maintaining full power during the cycle and can extend the operating cycle at rated power.
The O
magnitude f this extension is dependent on the characteristics of the core and on the maximum allowable core flow.
In general, operation with ICT can extend full power operation by approximately one week.
Final feedwater temperature reduction (F WTR) at the end of-cycle O
(EOC) can further extend the operating cycle.
In general, a 100 degree F reduction in feedwater temperature provides approxirmately two weeks of additional full thermal power operation.
The ICF region (Figure A.1-1) is referred t as the extended operating domain.
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O NEDC-31449 lQ A.2 PIR;T OPERATIONAL STATUS A.2.1 RECIRCU1ATION SYSTEM CAPABILITY O
Based on recirculation system performance data for QCNPS, the maximum core flow is expected to be somewhat less than 108% of rated core flow.
Given that the ICF capability results from a
system performance nsideration rather than a safety concern, the analyses are performed at O
the bounding condition of 108% of rated core flow.
A.2.2 MINIMUM END-OF-CYCLE COASTDO'iN Pok'ER LEVEL O
Standard licensing transient analyses are performed at the full power, end-of-cycle (EOC), all-rods-out condition.
Once an individual plant reaches this condition, it may shut down for refueling or it may be placed in a e asta wn m de f
peration. For ocNPS, this type of operation O
involves coasting down to a lower percent of rated power while maintaining constant core flow.
For the purpose of this safety evaluation, coastdown to 20% power is assumed.
O The power profile during this period ir as.sumed to be a linear function with respect to exposure.
It is expected that th+ actual profile will be a
slow exponential curve.
An analysis using the linear O
app m imati n, h wever, util be conservative, since it bounds the power level for any given exposure.
In Reference 3,
evaluations were made at l
90%, 80% and 70% power level points on the linear curve.
The results showed that the pressure margins from the limiting pressurization transient and the MCPR operating limits exhibit a larger margin for each of these O
points than the EOC full power, full flow case.
The LOCA analysis results O
A.2 1 0
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I f r the rated Power /10Si of rated core flow case is conservative for the
!O coastdown period since the power will be decreasing and ICF will be j
maintained.
Reference 4 presents the results of analyses to justify 4
4 coastdwn power operation.
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O NEoC-31449 A.3 SAFETY ANALYSIS O
A,3.1 ABNORMAL OPERATIONAL TRANSIENTS A.3.1.1 Limitine Transients
'O The limiting operational transients analyzed in the QCNPS L'ni t 1 Reload 9,
Cycle 10 supplemental reload licensing submittal (Reference 5) were re-evaluated for ICF followed by final feedwater temperature reduction (FF*'TR) es follows:
O Nuclear transient data for 100% power, 108% core flow (100P/10BF) with and without FFRTR (corresponding to a 100 degree F reduction in feedwater
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end of-cycle 10 (EOC 10).
These nuclear data were then used to analyze the load rejection without bypass (LRNSP), turbine trip without bypass (TTNBP) and the feedwater controller failure ( Fa*CF) events at the 100P/10SF condition.
The transient events were analyzed based on core characteristics with both BP/P8x8R and CE8x8EB.
The results of the transient analyses are presented in Tables A.3 1.
A. 3 2 an3 A. 3 3.
As shown in Tables A.3 2 and A.3 3, the limiting event (IRNBP) from Reference 5 bounds all of the t.CPR results from the ICT and ICF/FFw'TR analyses with one exception:
the BP/P8x8R limits for LANBP and TTNBP at ICF with rated feedwater temperature. Therefore, the current technical specification MCPR operating limits should be modified to incorporate the changes as shown in Table A.3-3 for operation at ICT conditions.
The transient performance responses for the 12NBP, TTSBP and Fa'CF events are shown in Figures A.3 1 through A.3 7 O
A.3.1.1 Overoressurination Analysis The limiting transient for the ASME code overpressurization analys
[nain steam isolation valve (MSIV) closure with flux scram (direct s c ra:t failure)), was evaluated for extended EOC10 conditions with ICF without i
A,3 1 iO
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FWTR (Table A.3 4 and Figure A.3 8).
For this evaluation, ICF without O
F WTR is more severe than with FWTR, and the MSIV closure with flux scram event provides the most limiting overpressure transient response.
The transient analysis (Table A.3 4) for the ICF condition produced a peak vessel pressure of 1327 psig, which is below the ASME Code upset limit of O
1375 psig and, therefore, is acceptable, A.3.1.3 Rod k'ithdrawal ErroI O
The rod withdrawal error (Rk'E ) transient was evaluated under ICF conditions.
k' hen ICT is e= ployed, the rod block monitor (REM) setpoint (which is flow biased) increases and gives a higher MCFR limi.
Thus, the REM should be clipped at flows greater than 100% of rated so that the ACPR
,O values determined without ICF (Reference 5) apply.
The clipping procedure includes an adj us t:re nt to the REM circuit so that the high RBM trip setpoint at flows greater than 100% of rated is equal to the value at 100%
rated fl v.
These results are independent of whether FWTR is inpleented O
or not, and are therefore bounding.
A.3.2 FUEL LOADING EP20R O
Operation with ICF and/or FWTR does not inpact the analysis of the rotated bundle fuel loading error event.
Thus, the results reported in the Cycle 10 reload licensing submittal (Reference 5) are applicable for operation with ICF and/or F WTR.
A.3.3 ROD DROP ACCIDEST The rod drop accident (RDA) event is a startup accident evaluated at O
cold and hot s ta ndb-; core conditions which are unaf f ected by ICF and FFTR operation.
Therefore, there is no change to the RDA analysis bases as presented in Reference 5,
and the RDA requirements of Reference 5 are applicable for operation with ICF and/or FWTR.
A.3-2
'O
O fiEDC 31449 A.3.4 LOSS-OF-COOLA!;T ACCIDE!?T (LOCA) A!;ALYSIS O
For core flows lower than a critical value, boiling transition at the limiting fuel node (i.e.,
the high power node) can occur sooner than observed at rated operating conditions.
This phenomenon is referred to as O
early boiling transition (EsT).
If EsT occurs for the high power node at reduced flow, the resultant peak cladding temperature (PCT) can exceed the rated condition results.
If there is no PCT margin to regulatory limits, it may be necessary to apply a maximum average planar linear heat O
generation rate (MAPLHGR) multiplier for operation in certain flow ranges.
Low flow effects were generically addressed in Reference 6,
which was approved by the Reference 7 !?RC Safety Evaluation Report.
It showed that no MAPLHCR multiplier is required for the QC!;PS class of plant.
O The effects of ICP and/or FFJTR on LOCA analyses are insignificant because the parameters which most s.trongly af fect the calculated PCT (i.e.,
high power node boiling transition time and core reflooding time) have been O
shown to be relatively insensitive to core flow and feedwater temperature chan6es of this magnitude. Both of these modes of operation tend to slightly improve the results.
Vith the lower initial core void fraction, there is more liquid mass to be lost out of the break before core uncovery O
results.
The net effect of void fraction and other effects will result in a LOCA PCT change of less than 10 degrees F, which is insignificant in view of the large PCT margins from the new SAFER /GESTR I.OCA analysis (Reference 8).
O Therefore, it is concluded that ths LOCA analysis results for QCf;PS are applicable and insensitive to operation with ICF and/or FFJTR.
10 3.3.3 7Htexst. HYDRAULIC STABILITY The core and channel hydrodynamic decay ratio were evaluated for ICF and/or FFJTR operation.
If the reactor initially operates at ICF and at or O
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'O
.O NEDC 31449 below the rated rod line, both channel and core decay ratio will be less
'O severe than for operation at rated core flow.
k'i th FSTR alone, the channel decay ratio would improve because of the increased subcooling effect but the core decay ratio could be slightly more severe.
Therefore, if only FnTR is utilized, operation should be at or below the rated rod
'O line.
However, the combined effect of operating the reactor with ICF first and then with FFb7R would result in a lower overall core and channel decay ratio than for normal operation.
- O Therefore, it is concluded that the reactor core stability and the channel hydrodynamic stability performance with ICF and/or FSTR are within the established criteria.
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Table A.3-1 CDRE-WIDE 'IRANSIDTP ANALYSIS RFEUL'IS AT ICF NID ICF/FFWIR Rated nwinn Maxian Feedwater Muimum Oore Avg.
Maxia n Steam
)
a b
mtp Neutmn Surface Vessel Line j
Redsp) ion Flux Heat Flux Press.
Press.
Trarsient Figurn Ebel Ibwer Flow
( F
(% NIR)
(% Initial)
(Ibia)
(Psia) thrcription FArirr E>mm
(% NIR)
(% NIR)
IRNBP Ref. 5 FDC 100 100 0
505.5 120.1 1230 1200 IRNBP A.3-1 EDC+128 100 108 0
517.7 120.3 1232 1198 IRNBP A.3-2 EDC+582 100 108 100 439.0 118.9 1214 1181 l
TINBP A.3-3 EOC 100 100 0
494.9 119.8 1229 1199
> TINBP A.3-4 00C+128 100 108 0
514.3 120.2 1231 1197 8i Oi
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u E TINBP A.3-5 EOC+582 100 108 100 440.7 117.9 1211 1178
- l
229.9 116.0 1138 1104 FWCF A.3-6 DOC +128 100 108 0
229.8 115.8 1140 1102 FWCF A.3-7 IDC+582 100 108 100 215.8 120.6 1120 1083 IRNBP = Ioad mjection with no bypass, TINBP = 'Ibrbine trip with no bypr.s a.
FWCF = Feedvater contmller failum at mxinn demrtl b.
= End-of-Cycle, DOC +128 = End-of-Cycle + 128 MWD /Mr EDC+582 = Dxl-of-Cycle + 582 MWD /Mr c.
Reduction in feed.nter terperature from nomiml rated feedvater terperature (340 F)
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q NEDC-31449 i
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Table A.3 2 4
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j CORE VIDE ACPR RESULTS Uncorrected Option A OptionJ Transients Exposure E/I F P.'TR BP/8XBR GE8X8EB BP/P8X8R GEBX8EB BP/P8X8R GEFX0EB O
LRNBP EOC 100/100 No 0.19 0.19 0.26 0.26 0.21 0.21 EOC + 128 100/108 No 0.20 0.19 0.27 0.26 0,22 0.21 EOC + 582 100/108 Yes 0.18 0.18 0.25 0.25 0.19 0.20 10 TTNEP EOC 100/100 No 0.19 0.19 0.25 0.26 0.20 0.21 EOC + 128 100/108 No 0.20 0.19 0.26 0.26 0.21 0.21 EOC + $82 100/108 Yes 0.17 0.17 0.23 0.23 0.18 0.18 O
l F.'CF EOC 100/100 No 0.14 0.14 0.20 0.20 0.15 0.15 EOC + 128 100/108 No 0.14 0.15 0.20 0.21 0.15 0.16 EOC + $82 100/108 Yes 0.17 0.17 0.23 0.23 0.18 0.18 l
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G NEDC 31449 Table A.3-3 O
MCFR OPERATING LIMITS FOR QCNPS UNIT 1, EOC 10 Option A Option B Transient BP/P9XcR GERXCEB BP/Ph>3 CE6XPEP I.F2:Er 1.33 1.33 1.28 1.28 (100P/100F)
LF2;LP 1.34 1.33 1.29 1 28 (100F/10EF, Rated Feedwater Terperature)
LF2:SP 1.32 1.32 1.26 1.27 (109P/10ST, Fra'IR )
v Table A.3 4 o
Q OVERFRESSUR12ATION A';ALYSIS RESULTS nu Maxiru-Maxirr Initial Initial Stearline Vessel Power Flow Pressure Pressure T M."S I E':T (6)
(6)
(PSIG)
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MSIV Closure - Flux Scrar 100 100 1295 1319 Reference 5 (Reference 5 EOC)
MSIV Closure - Flux Scrar 100 108 1290 1327 Figure A.3 8 O
(ICr v/o Fr.TR, EOC+12s L'D/MT )
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1 A.3-15 0
c
.O gggc.31449 A.4 MECHANICAL EVALUATION OF REACTOR INTERNALS AND FUEL ASSEMBLY
- O A.4.1 LOADS EVALUATIONS 6
Evaluations were performed to de te rrnine the bounding acoustic and l
O flow induced
- loads, reactor internal pressure difference
- loads, and fuel support loads for ICF and/or FWTR operation.
A postulated sudden break in the recirculation line is accompanied by
'O the propagation of a decompression wave which originates at the break and propagates back toward the vessel.
Once in the vessel, the wave would broaden and lose intensity.
However, it can create lateral loads on the vessel internals located opposite the recirculation suction line connections
.O to the vessel.
The pressure wave amplitude will be larger if the subcooling to the downcocer is increased and, therefore, the lateral loads could increase.
The high velocity flaw patterns in the downcomer resulting from a recirculation suction line break also create lateral loads on the reactor O
vessel internals.
rhese loads are proportional to the square of the critical mass flow rate out of tne break The additional subcooling in the downcorer resulting from FWTF operation can lead to an increase in the flow-induced loads.
The reactor internals rost impacted by acoustic and O
flow induced loads are the core shroud, shroud support and j e t pump s.
A reactF internals pressure difference (RIPD) analysis was perforned to evaluate the effect of ICT operation on the reactor internal components
.;O loadings.
The increased internal pressure differences across the reactor internals were computed for the 108% rated core flow at normal, upset and l
faulted conditions for the reactor internals impact evaluation.
O Eased on the results from plant specific fuel lift analyses perforrwd at 1084 core flow, the resulting impact of ICF operation on the fuel support loads and fuel isundle lift for QCNPS vere evaluated.
Fuel support loads and fuel bundle lift were evaluated for upset, faulted and fatigue load
'O combinations.
le was shown that the fuel bundle lift is expected to be l
A.4 1 O
l
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'O NEDC 31449 J
minimal, and the design basis vertical loads on the fuel assembly and its O
supports remain valid.
A,4,2 LOADS IMPACT O
A 4.2,1 Reactor Internals The reactor internals most affected by ICF and/or FFVTR operation are the core plate, shroud eupport, shroud, top guide, shroud head, steam dryer, control rod guide tube, control rod drive housing and jet purp.
These and other components were evaluated using the bounding pressure differential loads, as calculated in Section A.4.1, under normal, upset and faulted conditions. It is concluded that the stresser produced in these and other
.O components are within the allowable design lirrits given in the Final Safety Analysis Report or the ASME Code,Section III.
A.4.2.2 Fuel Asserblies The fuel assemblies, including fuel b.indl e s and channels, were evaluated for ICT operation considering the effects of loads discussed in I
Section A.4.1 under normal, upset and faulted load combinatiens.
Results of jO the evaluation demonstrate that the fuel assemblies are adequate to i
withstand ICF effects up to 10E4 core flow.
1 The fuel channels were also evaluated under normal, utset and faulted 10 conditions for ICT operation.
The channel wall prest.ure gradients were found to be within the allowable design limits.
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- Q NEDC 31449 i
A.5 rl.0k' INDUCED VIBRATION O
To ensure that the flow induced vibration response of the reactor inta zals is acceptable, a single reactor of each product line and size
_O una goes n extensiv, vibration test during initial plant sr.artup.
After analyzing the results of such test and assuring that all responses fall within acceptable limits of the established criteria, the reactor is classified as a valid prototype in accordance with Regulatory Guide 1,20.
O All other reactors of the same product lit.e and size undergo a less rigorous confirmatory test to assure similarity to the base test.
Both QCNPS Units 1 and 2 are Bk'R/3 251 inch size plants.
Reactor
,0 internal vibration nessureeents were conducted at ocSps Unit 1.
since Unit 2 internal co:rponents are similar to Unit 1, the sace approximate vibration levele are expected at Unit 2.
Test data from Unit 1 were analyzed based on current GE standard design bases procedures and acceptance criteria based on
'O a eaxirtu: allowable alternating stress intensity of 10,000 psi.
The results showed that, at the rated core flow condition (98 M1b/hr), the maximum vibration artplitude observed was 40% of the acceptance criteria.
In addition, the anp11tude of vibration is assumed to be prcportional to the lO square of the flow velocity.
Therefore, Icr operation at lost of rated core flow would increase the vibration level to apprcxicately 65% of the acceptance criteria.
Therefore, it is concluded that the vibration level recains acceptable for operation at 1084 core flow.
O O
O A.5 1 0
i O
A.6 FEED'JATER N0ZZLE FATIGUE ANA1.YSIS c0 An evaluation of the effe:t of rWTR and end of cycle coastdown on feedwater noz:le fatigue was performed for the QCNPS with the following assumptions:
- O (1) 3n 18.,tonth fuel cycle.
(2)
F WTR to 230 degrees F (equivalent to 100 degree F reduction) for 14 days was followed by a coastdown to 704 power over a O
period of 12 weeks.
The feedwater tenperature at the end of the coastdown was 210 degrees F.
A.6.1 METHODS AND ASS 1'MPTIONS
,O The fatigue experienced by the feedwater no:zle results from two phenomena: (1) system cycling and (2) rapid cycling.
System cycling is caused by maj or tenperature changes associated with system transients.
lO Rapid cycling is caused by snall, high frequency tenperature fluctuations
(
caused by mixing of relatively colder nozzle annulus water with the hot reactor water.
The colder water iepinging the no::le originates from l
leakage past the thernal sleeve secondary seal and from the boundary layer
- O of cold water formed by heat transfer through the thermal sleeve.
FWTR af fects only the rapid cycling f atigue usage for two reasons:
(1) the transient temperature variation associated with these rodes of
- O operation is sea 11 and thus does not affect the system cycling usage factor, and (2) the time spent at a reduced feedwater tenrerature is a significant contributor to rapid cycling fatigue usage.
An updated rapic l
cyc l '.ng analysis performed in Reference 10 was revised to include the
- O condition f or FWTR and coastdown to 706 power.
The feedwater duty cap. (Table 3 2 in Reference 10), was stadified to include the additional indices shown in Table A.6 1 for FWTR operation
'O with coastdown.
These additional indices model the coastdown as a l
l A.6 1 I
O
()
four step process.
The temperatures and flow rates are set for each step O
to give conservative results.
The percentage of time spent in the FFJTR and coastdown is subtracted from the percentage spent in normal operation.
Seal ring refurbishment time is determined so that by the end of the l
()
feedvater nozzle life, the sum of the system cycling and the rapid cycling usage factor for each the feedwater nozzle locations will not exceed the allowable value of 1.0.
It is assumed that the system cycling usage factor is linearly dependent on the number of years since the beginning of E) feedwater nozzle life.
Af ter each year, the total rapid cycling usage factor from the beginning of life is compared to the maximum allowable rapid cycling usage factor for each year, which is determined as follows:
O (1
UFMAX - Sctf. yrs I
LIFE
()
where:
U,,x
- Maximum allowable rapid c, cling usage factor from beginning of life for specific year O
UFMAX - Maximum allowable rapid plus system cycling usage fact.or by the end of the feedwater nozzle life f
()
SCUF - Total system cycling usage factor for the feedwater nozzle life i
LIFE - Feedwater nozzle design life (years)
C) i yrs
- Number of years since beginning of feedwater nozzle life 6
If the tstal rapid cycling usage factor since beginning of life exceeds U,,
for any feedwater nozzle locations, seal ring refurbishment is
()
A.6 2 0
O NEDC 31449 assured at the end of the previous yesr.
This method is illustrated in O
Figure A.6 1.
A.6.2 RESl'LTS The analysis docunented in Reference 10 indicated that re fufbi shn.ent of the thernal sleeve seals after 11 years would be necessary to keep the 40 year total fatigue usage (system cycling plus rapid cycling) below a value of 1.0.
Keeping the refurbishment schedule constant for the O
analysis, the 40 year total fatigue usage was calculated as shown in Table A.6 2.
The fatigue dar. age per cycle for FWTR operation is conservatively estinated by taking the difference between the FWTR fatigue and the normal operation fatigue and dividing that quantity by the number of cycles in 40 O
years.
If FWTR and coastdown were used for every cycle during the plant's life, the 40 year total fatigue usage factor would be greater than 1.0, O
assu=ing that the seals were replaced after 11 years.
Satisfactory fatigue usage can be achieved by reducing the refurbishnent interval to seven (7) years, as noted in Table A.6 2 assuming FWTR at the and of every cycle, the refurbishrent interval is inpacted by four (4) years.
O The results of this analysis are based on the expected coantdow.
operational strategry (Section A.6) and on leakage correlations developed during testing of the triple sleeve design.
A shorter end of cycle O
coastdown period or a smaller temperature reduction would increase the refurbishment interval.
Also, the leakage is based on several georetric f actors and ar.suned corrosion rates for the sleeve and safe end materials.
The resulting fatigue results are conservative for the expected plant O
operation mode.
Since leakage is the primary contributor to rapid cycling fatigue, a more accurate evaluation of rapid cycling could be made by monitoring seal leakage and considering actual plant performance.
O A.6 3 0
NEDC 31449 Table A.6 1 FEEDATER DUTY MAF INDICES ADDED FOR FFWTR AND COASTD0'.*N
)
Cycle Feedwater Feedwater Region A Time Index Flow Tergerature Tempgrature Year
()
(4 rated)
( F)
( F)
(%)
20 100 225 546 2.56 21 100 220 546 3.84 22 92.5 215 546 3.84 c3 23 85 210 546 3.84
~
24 77.5 205 546 3.84 O
Notes: (1) The faedwater terperature is based on a lower value of a +3n variation on the nominal terperature.
(2) The time spent at this node of operation (2.56 + 3.E4 + 3.84 + 3.84 + 3.84 - 17,924) was subtracted from normal operation (i.e., index 1 - 65.20 17.92 - 47.284).
()
O n
()
O O
A.6-4 O
C)
NEDC 31449 Table A.6 2 N0ZZLE FATIGUE USAGE FOR A 11 YEAR SEAL REFURBISHMENT PERIOD FOR FPJTR AND COASTDO*m'N (LOCATION I) i l
18 Honth Cycle
.)
FFb'TR/Coas tdown
(
Normal to 70% Power Operation Each Cye_1e 40 Year Total J[)
Fatigue Usage 0.5735 2.1843 Additional Usage Due To FF'a*TR and 1.6108 Coastdown
()
Additional Usage 0.0600 Fer Cycle
()
Note: The total 40 year usage factor for FPm*TR operation after every cycle can be kept to below 1.0 by refurbishing the seals after 7 years (at location D).
()
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Intervals A.6-6 O
9 hEDC 31449 A.7 C0!; TAI!;ME! I Af;ALYSIS The itopac t on the containment LOCA response was evaluated for QC!:PS with regard to operation in the expanded power / flow rtap (i.e.,
including q
operation above the rated rod line and in the ICT region with and/or without FWTR).
The irtpo r tant containrent pararreters cc.nsidered in the analysis av include:
(1) Dryvell pressure and temperature (2) Suppression charber airspace pressure and terperature 3
(3) Dryvell suppression chamber differential pressure (4) Suppression pool temperature (5) Annulus pressurination loads (6) Hydrodynamic loads.
O v
Results of the analysis showtu that the peak values of drvwell pressure and terperature, suppression chaeber airspace pressure and tenperature, suppression pool ter perature, and annulus pressurization leads qv are bounded by the values reported in the Plant Unique Load Definition (PULD) (Reference 11).
Maj or contain::ent hydrodynamic loads pos' ilate d to occur in a hypothetical LOCA vere evaluated and included pool swell load, condensation oscillation (CO), and chugging loads.
All these dynaric loads nO are boundec' by their corresponding design values except for the vent line thrust load.
The peak calculated value for the vent line thrust load is 21, higher O
than the PULD reported value.
This load represents only a part of the total rtaxirrum vent syster disenarge load (i.e the vent thrust Inad is just one corponent of the tr ax irtur vent discharg.e load cotbination).
Frcr QC::PS Plant Unique Analysis Report (Reference 12), the tra r g in for the O
reaxieue vent system discharge load (allowable versus calculated) is A.7-1 O
6 NEDC 31449 estirrated at 254.
The vent line thrust load contributes 194 to the total O
load corbination. Therefore, while this vent thrust load cor:ponent exceeds the PL'LD-va lue, the total vent discharge load te rr.a ins well within thc existing design n:argin.
O S/RV loads on the containtrent are not affect <
secause there is no change in the S/RV setroints or reactor operating pressure associated with operation in the extended operating dorain.
O O
O O
O O
O A.7 2 lO l
l 9
!;EDC 31449 4
PART B PIR;T EQUIPME!;T OUT OF SERVICE ANALYSIS FOR QUAD CITIES t. CLEAR PO'a'ER STATIO!; l'!;ITS 1 A!;D 2 T
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9 NEDC 31449 B.1 INTRODUCTION "e"
The purpose of this section is to present the results of a study prepared for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2 to establish the licensing bases for continued plant operation with a single failure of the following equipment:
(1) Last stage feedvater heater string (2) One safety / relief valve (S/RV)
(3) One recirculation purp loop The ability to operate at full power or at a reduced poser level throughout an entire or partial reactor fuel cycle with one of the above equipment out of service would be of significant economic value.
Operational flexibility and capacity factor are increased because the plant can continue to operate until the out of service equipment can be repaired U
or until the next convenient outage occurs.
issuning only a single equipnent failure, the resultant operating MCPR limits are applicable to the expande d power / flow nap (Figure A.1 1; with the exception of the increased core flow region.
O To establish the technical specification operating limits for each of the equiprnnt assumed out of service, one or more of the following concerns need to be addressed:
O (1) Core wide transient performance (2) Containment dynamie loads (3) Feedwater no::le fatigue n
(4) Loss of Coolant accident (LOCA)
O B.1-1 o;)
B.2 FEED'a*ATER HEATER OUT OF SERVICE l.,
This analysis justifies operation with 100 degree F reduction in feedvater tertperature in the expanded operating domain with the exception of the ICT region (Figure A.1 1).
The feedwater heater out of setvice analysis supports a contingency operating mode allowing continued operation 0
with reduced feedwater temperature over a full fuel cycle.
Operation with feedwater heater out of service is similar to operation wi t h FI'.'T R, except that (1) the duration of operation can be longer and,
'~
(2) operation can occur at any tice during the cycle.
Therefore, transient analyses are perforced to develop a cycle independent operating McrR limit applicable to plant operation at the reduced feedwater tenperature.
In addition, the irtpa c t on other safety analyses and design bares such as contaircent, LOCA and feedwater no::le fatigue is evaluated.
B.2.1 ABNORMAL TRANSIENTS Es'ALUATION o
Operating with a f e e d. cat e r heater out-of service could potentially irpact plant transient analysis as follows:
(1) The direct effect of reduced feedwater temperature is to
(')
increase the core inlet subcooling which in turn affects the core pressurization rate and reactivity during postulated transients C-)
(2) The potential change in core inlet conditions can affect the reactor nuclear paraneters such as the power shape and core void fraction, Changes in these paraneters can affect the plant responses fcir the transient events analyzed.
U, Te establish cycle independent operating lic.ts for reactor operaticn with a feedwater heater out-of service, a bounding end of cycle (EOC) exposure condition is used to develop nuclear input to the transient O,
B.2-1 0
GD NEDC-31449 analysis model.
The severity of the transient results is strongly dependent on the effectiveness of the control rod scram action.
For this reason, the EOC bounding exposure condition assanes a more top-peaked axial power distribution than the nominal power shape, thus yielding a poorer scram response.
Analyses with this bounding power shape result in a ACPR 0.04 worse than similar analyses with the nominal power shape
- and, J
therefore, should provide reasonable conse rvat isms for operating MCPR limits in future cycles.
e The limiting transient event from Reference 5,
[ load rejection without i) bypass (LRNBP)] was analyzed with the feedwater heater out-of-service.
With reduced feedwater temperature, the LRNBP will be less severe because of the reduced core steaming rate and lower initial void fraction.
The feedwater controller failure (FWCF) event, although not limiting in terms
')
of oCPR, has the potential to become more severe with a feedwater heater out-of-service and could become the limiting transient.
Therefore, both LRNBP and FWCF were reanalyzed with the bounding power shape at 100% core power /1001 core flow with a 100 degree F feedwater temperature reduction.
ga The results for the above transient analyses are presented in Table B. 2 - 1 and time histories of the key parameters are shown in Figures B.2-1 thrcugh B.2-4 Table B.2-2 presents the ACPR results for the events
- J analyzed.
To account for plant operation in the region above the rated rod line, the above transients were also evaluated at 100t power /87% flew and fe,und to be less limiting than the rated condition case.
As expected, the LRNBF eveni wiLL f J-utcr hcote; cu: Of service is less limiting than the
$)
base case with feedwater heater operable.
The opposite trend is observed for the FWCF event.
However, with feedwater heater out-of-service, both the LRNBP and FWCF event yield sity.ilar operating limits for this operating condition.
m v
Therefore, the operating limits associated with feedwater heater out -of-se rvice are :
C) 1.37 (Option A) and 1.32 (Option B)
B.2-2 C)
()
The above limits are valid for all future cycles at QCNPS Units 1 and 2 loading current GE fuel designs provided that:
O (1) The standard reload licensing LRNBP and TTNBP events result in operating limit MCFR (OLMCPR) values less than or equal to 1.37 and 1.32 for Option A and B, respectively.
O (2) The standard reload licensing analysis FWCF event results in sLMCPR less than 1.34 and 1.29 for Option A and Option B, respectively.
This condition is imposed to assure that the ACPR O
spread between the nCF and the LRNBP/TTNBP observed for the bounding power shape is maintained.
These two criteria are not expected to be restrictive, since they
()
represent conservative limits obtained with the bounding power shape.
The current Cycle 10 reload licensing analysis (Reference 5) results are also included in Table B.2.2 and confirm the 0.04 CPR margin establisned in the bounding basis.
O B.2.2 LOCA ANALYSIS Operation with a feedwater heater out-of service increases the
()
subcooling in the downconer and at the core inlet.
This could cause an increase in blowdown flow out of a postulated break in the recirculation line during the early stages of a LOCA.
This increase in subcooling and in blowdown flow cut of the break can cause several small effects on the ECCS 3
thermal-hydraulic analysis:
(1) The decay in core inlet flow could occur more rapidly because of the higher inventory loss and cause a slightly earlier fuel
()
cladding dryout (boiling transition).
(2) The core uncovery time could change slightly because of two competing effects: more mass loss out of the break, but more mass
()
in the core (due to lower initial core void fraction).
B.2-3 O
O NEDC-31449 (3) The sensible haat in the vessel and internals could be slightly I
lower which could affect the overall system depressurization
- O response.
The LOCA analysis of these effects demonstrated the insensitivity to changes in feedwater temperature of this magnitude.
The net in.pa c t of void fraction and other effects results in a LOCA PCT change of less than 10 E
degrees F, which is insignificant when compared to the conservatism in the standard LOCA analysis.
B.2.3 FEED'a'ATER N0ZZLE FATIGUE ANALYSIS An evaluation of the effect of a feedwater heater ou t - o f - s e rvic e on feedwater no :le fatigue was performed for QCNPS Units 1 and 2 with the M
following assumptions:
l (1) An 18-month fuel cycle.
l O
(2) Assuming a feedwater heater out of-service for various lengths of time at the and of a fuel cycle, which causes a 100 degree F drop in the feedweter temperature.
O A relationship was determined for increcental fatigue damage as a function of time spent at the lower feedwater tarperature.
As part of the e <pande d operating domain analysis (Section A.6),
a feedwater no :le fatigue study war perforced fot QCNPS operation with ICF/F.'TR.
Both O
IGF/FWTR ar.d f e e dw a'.e r heater out-of-service operation involre the same physical phenomena and fatigue mechanism to the feedwater no :le.
Therefore, the methods and assumptions previously described in Section A.6.1 remain applicable to the feedwater heater out-of-service condition.
O Also, the b asis for the results of this analysis are identical to the ICF/FF'.'TR analysis described in Section A.6.2.
Table B.2-3 shows the modification of the indices for feedwater heater O
out-of-service operation. The percentage of time spent in the feedwater B.2-4 O
(;)
heater out-of-service mode was subtracted from the percentage of time spent in normal operation.
A relationship of incremental fatigue damage as s O
function of time spent in the feedwat2r heater out-of-service mode was s
developed from these results.
The analysis documented in Reference 10 indicated that refurbishment iO of the thermal sleeve seals after 11 years is necessary to keep the 40 year l
total fatigue usage (system cycling plus rapid cycling) below a value of l
1.0.
Keeping the refurbishment schedule constant for the analysis, the 40-year total fatigue usage was calculated as shown in Table B.2-4 The O
fatigue damage per cycle for FrJTR operation is conservatively esticated by taking the difference between the FFJTR fatigue and the normal operation fatigue and dividing that quantity by the number of cycles in 40 years.
O If operation with a feedwater heater out-of-service were implemented for every cycle during the plant's life, the 40 year total fatigue usage factor would be greater than 1.0, assuming that the seals were replaced after every 11 years.
Satisfactory fatigue usage can be achieved by as noted in Table B.2-4.
O reducing the refurbishment interval to 8 years, The impact on seal refurbishment was less sevete for this rode of operation thav f or ICF/FFJTR.
.O B.2.4 CONTAINMENT LOADS EVALUATION The containment analys4.s results for ICF with F WTR (Section A.7) are applicable to the feedvater hecter out cf=cervice analysis nacausa che O
resulting core subcooling increase f ollowing ICF and FrJTR bounos che c r. 2 e for a feedwater heater out of service.
Given that FFJTR and feedwater heater out of service both result in a temperature reduction of 100 degrees F, the addition of ICF will increase the core inlet subcooling and yield O
conservative results if applied to the feedwater heater out-of-service analysis.
Therefore, the containment evaluation performed to support ICF with O
FFJTR is applicable to this feedwater heater out-of service analysis.
B.2 5 O
()
NEDC 31449 Table B.2-1 0
TRANSIENT ANALYSIS RESULTS FOR QCNPS AT 100P/100F FEEDWATER HEATER OUT OF-SERVICE O
Maximum Core Maximum Ave. Surface Maximum Maximum O
Transient Neutron Flux Heat Flux Dome Pressure Vessel Pres.
Description (5 NBR)
(S NBR)
(esir)
(psir)
LRNBP w/ FWH" 529.3 121.7 1193 1223 O
LRNBP w/o FWH 404.5 117.9 1174 1204 FWCF w/ FWH 267.8 118.0 1109 1141 FWCF w/o FWH 224.0 120.7 1086 1117
- O
^
FWH: feedwater heater 10 1
1 i
l O
O B.2 6 0
l l
lO NEDC-31449 Table B.2-2 O
OPERATING MCPR RESULTS FOR QCNPS AT 100P/100F FEED'.'ATER HEATER OUT OF-SERVICE
.O O
Exposure Transient 1 - Bounding Description 2 - EOC10
- ACPR, 01NCPR.
01MCP_B o
E LRNBP W/FWH 1
0.23 1.37 1.32 lO FvCF W/r.'H 1
0.18 1.31 1.26 l
IEBP W/0 nm 1
0.21 1.34 1.29 F.'CF W/0 FWH 1
0.21 1.34 1.29 O
LRNBP W/rm 2
0.19 1.33 1.28 F'CF W/ F.m 2
0.14 1.27 1.22
'O O
^
Uncorrected for Option A and Option B.
F.'H. feedwater heater.
O O
B.2 7 O
j
C)
NEDC-31449 Table B.2-4
()
N0ZZLE FATIGUE USAGE FOR A ll-YEAR SEAL REFURBISRMENT PERIOD FOR FEED'a'ATER HEATER OUT-OF-SERVICE (LOCATION I) 18 Month Cycle FVHOOS Operation
()
for 1314 Hours Normal Each Cycle Operation (10% per year) 40-Year Total Fatigue Usage 0.5735 1.5305 O
Additional Usage 0.9570 Due to FVH005 Additional Usage Per Cycle 0.0360 C)
Note: The total 40-year usage factor for 1314 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.99977e-4 months <br /> of
()
feedwater heater out of service operation during every cycle can be kept below 1.0 by refurbishing the seals after 8 years (at location D).
O O
O O
l B.2-9 C) l
[)
NEDC-31449 Table B.2-3
- O FEEDATER DUTY MAP INDICES ADDED FOR FEED'n'ATER HEATER OUT-OF-SERVICE Cycle Feedwater Feedwater Region A Time Year
!()
Index Flow Temperature Temgerature
(% rated)
( F)
( F)
(t) 20 100 225 546 10 O
Notes: (1) The feedwater teuperature is based on a lower value of a +3% variation on the nominal temperature.
(3 (2) One case was arbitrarily analyzed:
10% time per year - 1314 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.99977e-4 months <br /> / cycle.
l (3) The time spent at this mode of operation (10%) was subtracted from normal operation (i.e., index 1 - 65.20 55.20).
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C)
NEDC-31449 B.3 ONE SAFETY / RELIEF VALVE OUT-OF-SERVICE O
This analysis provides the technical bases for operation of QCNPS Units 1 and 2 with one safety / relief valve (S/RV) out-of-service.
In r) particular, the accident and transient considerations for operation with one S/RV out-of-service are presented.
B.3.1 ABNORMAL TRANSIENT EVALUATIONS O
Operation of QCNPS Units 1 and 2 with one S/RV out-of-service could affect the change in critical power ratio (ACPR) in the event of an abnormal operating transient.
The decrease in relief capacity could lead O
to higher pressures during a pressurization event, which could lead to a larger ACPR.
The failure of one S/RV could also result in a higher peak vessel pressure, thereby reducing the margin to the ASME upset code limit for a pressure vessel.
O The transients which yields the most limiting ACPR for QCNPS Unit 1 Cycle 10 is the load rejection without bypass (LRNBP).
This event was reanalyzed with the most limiting relief valve disabled (i.e.,
the relief
()
valve with the lowest setpoint).
The valve setpoints used in this analysis are given in Table B.3-1.
For the overpressure criteria, the main steamline isolation valve (MSIV) closure transient with high flux scram was analyzed with the lowest setpoint spring safety valve out-of-service.
'O B.3.1.1 Impact on Delta CPP. Analvsis The LRNBP transient was analyzed using the ODYN computer program with full relief capacity (as a base case) and with the lowest setpoint relief n
!O l
E.3 1 lO
O NEDC-31449 l
valve out-of-service. The results showed no change in the oCPR due to the lO reduced relief capacity.
Plots of typical transient responses are shown in Figures B.3-1 and B.3 2.
From the transient responses, it can be seen that the peak neutron O
flux occurs about 0.7 second before the relief valves open for both l
analyzed cases.
Because QCNPS has two relief valves with the same low setpoint, disabling one relief valve affects only the pressure relief capacity and not the time of valve initiation.
Because the neutron flux is O
decreasing rapidly at the time when the relief valves open, a change in the overall relief capacity will not affect the CPR result.
In summary, with one relief valve out of-service there is negligible O
impact on the MCPR limit.
The oCPR for this operating condition will be bounded by reload licensing calculations.
This conclusion is valid for current General Electric fuel types and analysis methods as applied to QCNPS Units 1 and 2.
O B.3.1.2 Irreac t on overrress re Criteria O
Reference 9 documents the results of cE sensitivity studies which show the effect of a S/RV out of-service is a peak pressure increase of less than 20 psi.
O The adequacy of the S/RV capacity based on ASME code requirements is demonstrated by the MSIV clos;re transient with high flux scram and without credit for relief valve operation.
With the lowest setpoint spring safety valve out of-service, this transient event still shows an adequate margin O
of 54 psi to the ASME upset code limit of 1375 psig.
The time response of key variables for this transient is shown in Figure B.3 3.
O l
1 B.3 2 l
0
O NEDC-31449 B.3.2 LOCA ANALYSIS O
If the out-of-service valve has an automatic depressurization function (ADS), there can be potential impact on the calculated peak cladding o
2 temperature (PCT) for small break sizes of less than approximately 0.2 ft O
With a worst case postulated single failure of the High Pressure Core Injection (HPCI) System, a small effect may be seen because the small break transient is dominated by the time required to depressurize the reactor to the operating pressure of the low pressure Emergency Core Cooling System O
(ECCS).
The effect of one ADS valve out of-service was accounted for in the Reference 8 LOCA analysis because only four of the five ADS valves were O
used for the break spectrum analysis.
The results showed much lower PCT values for small breaks than for the large breaks, which are the most limiting LOCA cases for this plant.
For the large breaks cases, the reactor vessel is rapidly depressurized prior to the actuation of the ADS.
O Therefore, cine ADS valve out of-service has no impact on the calculated PCT.
O O
l O
1 O
B.3 3 O
3 NEDC-31449 Table B.3-1 VALVE SETPOINTS USED FOR TRANSIENT ANALYSIS lO Setroint (e s i rl Type No.
.O 1105 + 1%
RV 1
1125 + lt RV 2
1125 + lt S/RV 1 O
O O
- One S/RV assumed out-of service lO O
O B.'
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isiw a towse B.3-7 0
O NEDC-31449 B.4 ONE RECIRCULATION PUMP OUT-OF-SERVICE O
From a plant availability / outage planning standpoint, the capability O
of operating at reduced power with a single recirculation loop is highly desirable in the event that traintenance of a recirculation pump or other components render one loop inoperable.
To justify the single-loop operation (SLO), accfdents and abnormal operational transients associated O
with power operation were reviewed for the single loop case with one pump in operation.
This SLO analysis was previously performed for QCOPS Units 1 and 2 and documented in Reference 2.
O To support the introduction of the GE8x8EB fuel design and the additional operating domain above the rated rod line (Figure A.1-1), the issues addressed by the referenced SLO analysis are reviewed to ensure their applicabilities with these operational changes.
In addition, the O
ittpact on safety limits for Sto in the region above the rated rod line with one safety / relief valve (S/RV) out-of-service is also addressed here.
B.4.1 MINIMUM CPR FUEL CIADDING INTEGRITY SAFETY LIMIT O
Except for the total core flow and the traversir.g in-core probe (TIP) measurements, the uncertaincies used in the statistical analyses to determine the fuel cladding integrity MCPR safety limit do not depend on O
whether coolant flow is provided by one or two recirculation pumps.
Since the core flow uncertainty and the TIP noise uncertainty are not affected by the proposed operational changes (i.e.,
the GE8x8EB fuel design O
and operation above the rated rod line with or without one S/RV out-of-service), the conclusions shown in the referenced SLO analysis are still applicable.
O B.4 1 O
I O
NEDC-31449 B.4.2 MINIMUM CPR OPERATING LIMIT O
The referenced SI4 analysis (Reference 2, Paragraph 6.B.3) demonstrated that, within the normal operating domain, the consequences of abnormal operation transients from one-loop operation will be considerably less O
severe than those analyzed for a two loop operation mede.
Operation with one recirculation loop results in a maximum power output significantly below (by 20 to 309.) that which is attainable with a two-pump operation.
Thus, for pressurization, flow decrease and cold water increase transients, O
the results for two-pump operation cases bound both the thermal and overpressure consequences of one loop operation.
The introduction of GE6x8EB fuel in the core is not expected to alter the above conclusion.
The observed transient performance trend (one-pump case bounded by two-pump O
case) remains applicable to the QCNPS cores with the new GE8x8EB fuel design.
The failure of the S/RY with the lowest setpoint for two-pump O
operation was previously shown to have no impact on the MCPR operating limits and the vessel overpressure criteria (Section B.3).
For SLO, the same conclusion remains applicable because the peak neutron flux would still occur before any S/RV actuation.
In addition, the lower initial O
power level for SLO mode would reduce the severity of the vessel peak pressure ccmpared with the two-pump case.
The above conclusions are also applicable for plant operation in the O
region above the rated rod line.
B.4.3 STABILITY ANALYSIS O
The introduction of GE8x8EB fuel design to QCNPS Units 1 and 2 cores will result in an insignificant impact on the core and channel decay ratio for reactor operation with one recirculation loop.
Therefore, the conclusions stated in the Reference 2 regarding this subj ec t are still O
applicable to QCSPS Units 1 and 2.
B.4 2
{
O
[?
NEDC-31449 B.4.4 LOCA ANALYSES O
The LOCA analysis documented in Reference 2, Paragraph 6.B.6, imposed a MAPLHGR reduction factor of 0.84 to GE 8x8 retrofit fuel type.
O Based on analysis experiences in using the SAFER /GESTR LOCA evaluation models, the GE8x8EB fuel design has been shown to have larger margins to the PCT limit than the 8x6R and BP/P8x8R fuel types. This is primarily due to the decrease in initial stored energy of the GE8x8EB fuel attributable
?)
to the increased initial pressurization level.
The Reference 8 analysis concluded, with SAFER /GESTR. no MAPuiGR multiplier is required for SLO at QCSPS Units 1 and 2.
O O
O O
O 10 I
B.4 3 0
()
HEDC 31449 REFERENCES O
1.
"General Electric Boiling Water Reactor Extended Load Line Limit Analysis for Quad Cities Nuclear Power Station Unit 1 Cycle 7 and Unit 2 Cycle 6", General Electric Company, July 1982 (NEDO-22192).
O 2.
"Dresden Nuclear Power Station Units 2 and 3 and Quad Cities Nuclear l
l Power Station Units 1 and 2 Single Loop Operation", General Electric Company, December 1980 (NEDO-24807).
O 3.
R.A. Bolger, Commonwealth Edison Co., Letter to B.C. Rusche, Director of Nuclear Reactor Regulation, USNRC, "QC-2 Proposed Amendment to Facility License No. DPR 30, Docket No. 50 265", June 11, 1976.
O 4
R.E. Engel, General Electric Company, Letter to T.A. Ippolito, USNRC, "End of Cycle Coastdown Analyzed with ODYN/TASC", September 1, 1981.
()
5.
"General Electric Boiling Vater Reactor Supplemental Reload Licensing Submittal f.r Quad Cities Nuclear Power Station Unit 1 Reload 9",
General Electric Company, June 1987.
()
6.
R.L. Cridley, General Electric Company, Letter to D G Eisenhut, USNRC, "Review of Lew Core Flow Effects on LOCA Analysis for Operating BWR s", May 8, 1978.
7.
D.G. Eisenhut, USNRC, Letter to R.L. Gridley, Genercl Electric Company, enclosing "Safety Evaluation Report Revision of Previously Icposed MAPLRGR (ECCS LOCA) Restrictions for BWRs at Less Than Rated Core Flow", May 19 1978.
C)
I 8.
"Quad Cities Nuclear Power Station Units 1 and 2 SAFER /GESTR Loss-of-Coolant Accident Analysis",
General Electric Company, June 1987 (NEDC 31345P).
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O NEDC-31449 9.
"General Electric Standard Application for Reactor Fuel", General O
Electric Company, May 1986 (NEDE-240ll-P-A-8-US).
10.
G.L. Stevens, B.J. Cheek, "Economic Generation Control Fatigue Usage Evaluation for Dresden Units 2 and 3 and Quad Cities Unit 1 and 2',
O General Electric Company, August 1984 (AE-78-0884).
11.
"Mark 1 Containment Program Plant-Unique Load Definition Report Quad Cities Station Units 1 and 2", General Electric Company, April 1982 O
(NtDO.24ss7 Rev.2).
12.
"Quad Cities Units 1 and 2 Plant Unique Analysis Report", Nutech Report No. COM-02 039, May 1983.
O 13.
"ARTS Improvement Program Analysis for Quad Cities Nuclear Power Station Units 1 and 2", General Electric Company, June 1987 (NEDC 31448P).
O O
ACYJ:0WLEDGEMENTS The analyses included in this report were performed by the combined O
efforts of many individual contributors, including:
P.F. Billig, T.G. Dunlap, G.D. Galloway, O
M.O. t.en:, L.x. 1.iu, H.X. Nghim,
G.L. Stevens, T.P. Shannon, J. Wallach and x. Jarhanian.
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