ML20247H841
| ML20247H841 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/31/1989 |
| From: | Charnley J, Lambert P, Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19292J340 | List: |
| References | |
| 23A5896, 23A5896-R01, 23A5896-R1, NUDOCS 8907310151 | |
| Download: ML20247H841 (44) | |
Text
-_- _ _ _.
_.t-23A5896 REVISION 1 CLASS I
) ~
MARCH 1989
)-
(23A5896, REV.'1)
)
SUPPLEMENTAL RELOAD LICENSING REPORT FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 1
)
RE14AD 6 CYCLE 7 (WITHOUT RECIRCULATION PUMP TRIP)
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Prepared:
4 'rA 0
P. A. Lambert Fuel Licensing Verified:
[FuelLicensing L. Rash
)
Approved:
/
4&
J.. S. 'Cfiarnfey, Mahager
~
Fuel Licensing
)
9 1
GENuclearEnemy y
175 Curner Annue SonJost. CA 95125 0
1 8907310151 890725
)
PDR ADOCK 05000324
'P PNU o_ __
23A5896' Rev. 1 IMPORTANT NOTICE REGAADING CONTENTS OF THIS REPORT
)
PLEASE READ CAREFULLY This report tas prepared by the General Electric Company (GE) solely.for Carolina Power and Light Company (CP&L) for CP&L's use with the United States Nuclear Regulatory Commission (USNRC) for amending CP&L's operating license for the Brunswick Steam Electric Plant Unit 1.
The information contained in this report is believed by GE to be an
)
accurate and true representation of the facts known, obteined or provided to GE at the time this report was prepared.
The only undertakings of GE respecting information'in this document
)~
are contained in the Contract between Carolina Power and Light Company and General Electric Company for Reload Fuel Supply and Related Services for Brunswick Steam Electric Plant Unit 1, effective December 31, 1982, as amended, and nothing contained in this document shall be construed as
)
changing said contract. The use of this information except as defined by said Contract, for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither GE nor any of the contributors to this document makes any
)
representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document er 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 which may result for such use of such information.
L
)'
2
23A5896 Rsv. 1 ACKNOWLEDGEMENT The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Report were performed by R. E. Polomik and T. P. Lung of the Fuel Engineering Seetion.
)
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)
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3
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.23A5896i Rev. 1 1.
PLANT-UNIOUE' ITEMS (1.0)*
Appendix A:
Limiting Conditions for Operation Appendix B:
Bases for Limiting Conditions for Operation Appendix C:
Plant Parameter Differences Appendix D:
Use of CEXL-PLUS Methods for Cycle.7
)
Appendix E:
Use of New Safety Limit MCPR for Cycle 7 LAppendix F:
Main Steamline Isolation Valve Out-of-Service
.2.
REIAAD FUEL BUNDLES (1.0. 2.0. 3.3.1 AND 4.0)
):
Fuel Tyne Cvele Loaded' Number 1rradiated P8DRB284H' 4
8 P8DRB299 4
8
)-
BP8DRB299 5
1E4 BP8DRB299 6
176 New
)
BD339A 7
60 BD323B 7
12h Total 560
)J 3.
REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average' exposure at
'end of cycle:
21,072 mwd /MT Minimum previous cycle core average exposure 1
at end of cycle from cold shutdown considerations:
20,481 mwd /MT Assumed reload cycle core average exposure at end-of cycle:
21,230 mwd /MT b
Core loading pattern:
Figure 1
)
- (-) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. A letter "S" preceding the number refers to the U.S. Supplement, NEDE-24011-P-A-8-US, May 1986.
4
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'l 23A5896 Rsv.-1 i
4
~ CALCUIATED CORE EFFECTIVE ' MULTIPLICATION AND CONTROL SYSTEM
. WORTH - NO VOIDS. 20*C (3.3.2.1.1 and 3.3.2.1.2) i Beginning of_ Cycle, K,ff Uncontrolled 1.111 Fully Controlled 0.966 L
. Strongest Control Rod Out 0.988 R. Maximum increase in Cold Core Reactivity 0.002 with Exposure into Cycle. AK
)
5.
STANDBY LIOUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (AK)
REE (20*C. Xenon Free) 600~
0.036
)'
6.
RRIDAD-UNIOUE TRANSIENT ANALYSIS INPUT (3. 3. 2.1. 5 AND S. 2. 21 (Cold Water injection Events Only)-
).
Void Fraction'(%)
41.7 Average Fuel Temperature (*F) 1096-Void Coefficient N/A* (g/% Rg)
-6.710/-8.387
)l Doppler Coefficient N/A* (g/*F)
-0 201/ 0.191 Scram Worth N/A* ($)
)
)
- N
. Nuclear input Data, A - Used.in Transient Analysis
)
- Ceneric exposure independent values are used as given in " General Electric Standard Application for Reactor Fuel." NEDE-24011-P-A-8 dated May 1986.
5
)
23A5896.
~Rav. 1.
7.
DRinAD-UNIOUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAKETERS (S.2.2)
Bundle i
Fuel
'Peakina Faetors Power Bundle Flow Initial-Dggign lE;Al Radial 63131 R-Factor (MWt)
(1000 lb/hr) (MCPR)
Exposure:
BOC7 to E007-2000 mwd /ST BP/P8x8R 1.20 1.51 1.40 '1.051 6.409
'109.5 1.25-1 L
CE8X8EB-1.20 1.51 1.40 1.051 6.418 112.2 1.26
- Exposure:
E007-2000 mwd /ST to EOC7 BP/P8x8R 1.20
'1.44 1.40' 1.051 5.127 111.7 1.31 GE8x8EB 1.20 1.45 1.40 1.051-6.139 114.3 1.32
)'
8.
SFYICTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization No
)
Recirculation Pump Trip:
No-Rod Withdrawal Limiter:
No Thermal Power Monitor:
Yes Measured Scram Time:
No f-Exposure Dependent Limits:
Yes Exposure Points Analyzed:
EOC7-2000 mwd /ST and EOC7 9.
OPERATING FLEXIBILITY OPTIONS (S.2.2.3) y Single-Loop Operation:
Yes Imad Line Limit:
Yes Extended load Line Limit:
No
)'
Increased Core Flow:
No Flow Point Analyzed:
N/A Feedwater Temperature Reduction:
No f
ARTS Program:
No Maxiaum Extended Operation Domain:
No
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6
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23A5896 Rev. I 10.
CORE-VIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
Methods Used: GEMINI Exposure Range: BOC7 to EOC7 Flux Q/A ACPR Transient
(% NBR) (% NBR) EP/P8x8R GE8x8EB Firure
)-
Inadvertent HPCI 122 119 0.15 0.15 2
Exposure Range:
BOC7 to EOC7-2000 mwd /ST
)
Flux Q/A ACPR Trangient
(% NBR) (% NBR) isP/Psx6R GE8x8EB Ficure Load Rejection Without Bypass 482 120 0.18 0.18 3
Feedwater Controller Failure 285 116 0.12 0.13 4
)
Exposure Range: EOC7-2000 mwd /ST to EOC7 Flux Q/A ACPR Transient
(% NBR) (% F1.R1 BP/P8x8R GE8x8EB Firure
)
Ioad Rejection Without Bypass 592 125 0.24 0.25 5
Feedwater Controller Failure 418 121 0.18 0.19 6
)
11.
LOCAL ROD WITHDRAVAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(S.2.2.1)
Limiting Rod Pattern:
Figure 7 Rod Block Rod Position ACPR
)
Readine (%)
(Feet Withdrawn)
BP/P8x8R GE8x8EB 104 3.5 0.11 0.11 105 3.5 0.11 0.11 106 4.0 0.13 0.13 107 4.5 0.14 0.14
)
108 5.5 0.18 0.18 109 12.0 0.21 0.21 110 12.0 0.21 0.21 Setpoint Selected:
107
)
7 1
)
23A5896 Rev. 1
'12.
CYCLE MCPR VALUES (S.2.21 Non-Pressurization Events
' Exposure' Range:
Inadvertent HPCI 1.19 1.19 L
1.25 Fuel Loading Error Rod Withdrawal Error 1.18 1.18 k
Pressurization Events Ootion A Ootion B-I BP/P8x8R GE8x8EB BP/P8x8R GE8x8EB Exposure Range:
1 BOC7 to EOC7-2000 mwd /ST Load Rejection Without Bypass 1.32 1.32 1.25 1.25 Feedwater Controller Failure 1.22 1.23 1.20 1.21
)..
Exposure Range:
.EOC7-2000 mwd /ST to EOC7 Load Rejection Without Bypass 1.34 1.34 1.30 1.30 Feedwater Controller Failure 1.27 1.27 1.24 1.24 13.. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3)'
P P
)
si v
. Transient (esic)
(esic)
Plant Response MSIV Closure 1235 1266 Figure 8 (Flux Scram) l 8
23A5896 Rev. 1 14.
LQ& PING ERROR RESULTS (S.2.5.4)
Variable Water Gap Disoriented Bundle Analysis: Yes*
Event A.QlB Misorientad 0.19
)
15.
CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1) j Bounding Analysis Results:
Doppler Reactivity Coefficient:
Figure 9
)
Accident Reactivity Shape Functions:
Figures 10 and 11 Scram Reactivity Functions:
Figures 12 and 13 Planc-Specific Analysis Results:
)
Resultant Peak Enthalpy, Cold:
149.6 Resultant Peak Enthalpy, HSB:
215.6 16.
STABILITY ANALYSIS RESULTS (S.2.4)
)
GE SIL380 recommendations have been included in the Erunswick Steam Electric Plant Unit 1 operating procedures and/or Technical Specifications and, therefore, the stability analysis is not required.
)
NRC approval for deletion of a cycle-specific stability analysis is documented in Amendment 8 to NEDE-240ll-P-A-8-US.
)
)
)
- ACPR penalty of 0.02 for the tilted disoriented bundle is applied to the cycle MCPR value reported in Section 12.
9
)
23A5896-Rev. 1
- 17. IDSS-OF-COO 1 ANT ACCIDENT RESULTS (S.2'. 5.2) j i
LOCA Method Used:
SAFE /REFLOOD/ CHASTE T-l
" Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit No.
1," General-Electric Company (NEDO-24165, December 1978, as amended and NEDE-24165-P, April 1988.)
1.
Fuel Tvoe: BD323B (CE8x8EB) 7 Average MAPLHGR (kW/ft)
Planar Exposure Most Least.
Oxidation (GWd/St)
Limiting Limitine PCT (*F)
Fraction
?.
0.0 11.29 11.73 1993 0.04 1.0 11.44
'11.85 2006 0.04 2.0 11.61 11.99 3.0 11.81 12.15 4.0 11.99 12.33 5.0 12.17 12.52 2111 0.06
)
10.0 13.10 13.10 2197 0.08 12.5 12.95 13.09 2189 0.08 15.0 12.75 12.78 2152 0.07 20.0 12.07 12.08 2062 0.05 25.0 11.37 11.40 2006 0.04 45.0 8.52 8.60 1684 0.01
)-
50.0 6.06 6.14 1581
<0.01 2.
Fuel Tyve: BD339A (CE8x8EB)
).
Average MAPLEGR (kW/ft)
Planar Exposure Most Least Oxidation (GUd/St)
Limitine Limitine PCT (*F)
Fraction 0.0 10.91 11.33 1993 0.04
)
1.0 11.16 11.54 2006 0.04 2.0
-11.43 11.75 4.0 11.82 12.18 5.0 12.02 12.30 2111 0.06 10.0 12.88 12.90 2197 0.08 12.5 12.84 12.85 2189 0.08 J'
15.0 12.59 12.59 2152 0.07 25.0 11.58 11.59 2006 0.04 35.0 10.36 10.38 1851 0.02 45.0 8.79 8.87 1684 0.01 50.0 6.22 6.31 1581
<0.01
)
10
)
i l
23A5896-Rsv. 1 i
.MMMMM.
.MMMMMMMMM.
MMMMMMMMMMM
- .MMMMMMMMMMM.
'::MMMMMMMMMMMMM r
: M M M M M M M M M M M M M
- M M M M M M M M M M M M M
- M M M M M M M M M M M M M
'::MMMMMMMMMMMMM
- - "MMMMMMMMMMM" l
MMMMMMMMMMM l
"MMMMMMMMM" "MMMMM" l I I l 1 I I l-1 I 1 357 9111315171921232527293133353739414345474951
)
FUEL TYPE A - P8DRB284H D - BP8DRB299 (Cycle 6)
B - P8DRB299 E - BD339A C_- BP8DRB299 (Cycle 5) F - BD323B l
)
Figure 1.
Reference Core Loading Pattern 11
)
23A5896 Rev. 1
\\l 1
l I
)'..,
a 1 NEUIRON FLUX
{ VE$5EL PRf$$ R!$((PS])
Avt SLptF ACLW At f t l'Y 2 REL lEF VALvt FLOW IS O. ?
/
'i 333,4 j
l
.4-
^
r" i
~
\\
io...
r
{
b I....
r
.. n..
- 0. 8 58.8 188.8
- 0. 0 58.0 800.6
)
71ME (Eco@S) t!Mt (Ecoes)
I Ltv :L(INCH.REF.S[P.SKRT) a v01 ) RE ACTIVITY 2 VESiEL STEAMFLOW 2 00P'LER Rf ACT :VITY
)
15 s.0 1.0
~
4 _.
./..~..
)
y.
-c
?.
E.i..
t
)
.s. s s e. e i n..
flME(KCOW53 TIME tKcc e$)
)
Fi ure 2.
Plant Response to Inadvertent Activation of HPCI 5
)
i 12
)
23A5896 Rev. 1
[.
I NEUTRON FLU <
1 VESSEL PR;15 RISE (PSI) 2 AVE $URFACE >E AT FLUX 2 5AFETY VALW: FLOW bh 150.0 398.0 I
8 2.s.,
le s. e -
h t....
(
N N
.. e e.e
- 8. 0 2.8 4.0 8.0 8.9
- 2. 0 4.4 5.0
)
TIME ( ECO @ $1 flME ( E COsos) 1 LEVEL (INCH-4EF.EEP-SKRT)
! v01D REACTIvlTY iMh5Ea
!. E,0.S.. '."J.u,iM...."
).
~
s E...
/
l
~U/
3 T
Qr y
e C
L;;
i....
I V
)
-i....
o
- i.,
TIME t.ECO@S)
TIME ESECO@$)
)
i
,i Figure 3.
Plant Response to Generator Load Rejection Without Bypass
)
(EOC7-2000 mwd /ST)
I i
l 13
)
i l
l i
23A5896 Rav. 1
)
150.0 1 NEUrRON fl X
1 YC$ iEL PRE $$ R!$E(PSI) 2 AVE SURFU HEA FLUX 2 $AF :TY WALVF FLOW 3 CDP [ INLI FLOW 3 REL lCF VALV 15 0.0 I
4 eYP($$ WAL
)-
M A
- d
-n b
l se.s o
50.9 I
- +
)
N i
- c: : :: :: '=
8.0
- 8. 0 18.0 20.9
- 6. 9 18.8 29.0
)
TIME (KCoes) fint (KCoes) i LEV:Lt!NCH-REF.$EP.SKAT) 8 V01 ) RE AdjVITY 2 YESAL STEAMFLOW 2 DOP'LER 8E%CT:v!TV a gga:nage gl3 !!
3ygggp5Agtov
)
33..,
\\
^ - " "
8.8 los.o
\\/
)
E>
$ 0. 0
\\
h.l.s 7
5 W
)
i
..0
- e. o l e.
r e.
- e. :
i s. :
re. e TIME 89ECO @ $)
fly (SECD@5)
).
Figure 4.
Plant Response to Feedwater Controller Failure
)
(EOC7 2000 mwd /ST)
I' 3
33A5896 Rev. 1
)
estuinaw rtus a vEsstL Petens misttPst) 2 Avf s M ACE 6t AT FLUX 2 SAFETY WALWL Flow a cwt isn rLov
!!!gg gAtp gg
,, s. e
)
G i.e..
)
ese..
a Y
3...
)
'\\
9 e.e e.e e.e s.e e.e
- s. e e.e s.e
<.e s.e j
TIM [ ( K CoeOS)
Tint (Etsues) i
)
I LCVELt!NC 1Cr-SCP.CKRT) 3 WD1D REACT!v1TY 2 Vt&SEL STEATLDW 2 IKPPLER IIL A:T!vlTY l\\
! E*" 5^c!!!!"
!!M8EPP1
i.e
)
,e e. e
'V
[e.e.~.1
/
, e,. e
{ W
[
_/
w
)
e t
8
... {
f
)
. is e.
s..
e..
s..
e.e s..
flut (KtseCS) fint ( KCSICS)
Figure 5.
Plant Response to Generator Load Rejection Without Bypass
)
(EOC7) 15
)
1 1
23A5896 Rov. 1 1
l 1
)
I NEUIRON bdM 3 WES!EL PRES!
2 AVE SURFA:t EA FLUX 2 SAT :tv V Alvi' RISE (PSI)
Lov
!cf 15!; M'
!$!!! ;^t;ii 's n..
w t
)
.L-n E
io...
b gr.
V c~,
: c; : :;
=: :
L n...
trat :Eco@si vinc sucmos
)
IIsM'E?IJl[55'"S*"
$EUS^EE Elivity
?mu s n a s o-fea nedmn
)
, - - i; -y_
4 E
I s.. e s
\\
N-s.e i
W
)
- e. e i s. e 2 3..
..e i s. e n..
flME ( E coe s) 11ME ESEco@$3
)
Figure 6.
Plant Response to Feedwater Controller F;11ure
)
(EOC7) 16
)
23A5896 Rsv. 1 V
2 6
10 14 18 22 26 30 34 38 42 46 50
)
51 36 36 47 6
6 6
6 6
43 36 36 36 36 36 36
)'
39 6
6 14 6
6-35 36 36 36 36 36 36 31 6
6 14 0
14 6
6
)
27 36 36 44 44 36 36 23 6
6 14 0
14 6
6 19 36 36 36 36 36 36
)
15 6
6 14 6
6 11 36 36 36 36 36 36 7
6 6
6 6
6
)
3 36 36 NOTES:
1.
No. indicates number of notches withdrawn out of 48.
Blank is
)
a withdrawn rod.
1 2.
Error rod is (26,31),
i
)
Figure 7.
Limiting Rod Pattern for Rod Withdrawal Error 1
17
)
23A5896 Rzv. 1 i
l
)
l l
3 NEUTRON F UX 1 VESSEL PMESS RISE (PSI)
I 2 Avt SURF A:E HEAT FLUX 2 SAFEf t VA.vt FLOW L
!b b
15...
3...a
~ -
v E 3....
N l
4'.
g l-W S e.
- i....
)
(.
_ J _. _
8.9 5.0 0.0 5.9 TIME isECONOS3 T!stt ISEcceS)
)
1 LEVELtINC4 REF.SEP.GKRT) 1 ID REACTIVITY i..
2.c,a., 9E..!.?.R..i.4.."
!. ?.M,h. 55!,7, 3
).
. _ ~.
~
a E.....
9
)
w A.
~
E
.m "N.
i,...
)
.i....
S..
Tint (stC0es)
TIME (SEco@S)
)
)
Figure 8.
Plant Response to MSIV Closure (Flux Scram) 18
)
23A5896 Rav. 1 I
)
0.0
)
i
-5.0
?-
to Y -10.0 g
g Z
)'
[ -15.0
(
v La. -
b-O g -20.0 o
)
-25.0 I
A CALCtLATED VALUE-COLD B CALCLLATED VALUE-SB C BOUNE VAL 280 CAL G COLD
)
D BOUNE VAL 280 CAL G HSB
-30.0 0.0 500.0 1000,0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.
)
l l
)
Figure 9.
Fuel Doppler Coefficient in 1/a*C l
l I'
).
23A5896 Rev. 1 L-20.0 t
17.5
).
15.0 m
g
/
=
=
12.5 x.
4d O
10.0
).
D D
]
-3 7.5 i
5.0
)
2.5 A ACCIDENT l' UNCTION B BOUNDING uALUE280 CAL /G 0.0d.0 5.0 10.0 15.0 20.0
)_
ROD POSITION, FEET OUT l
l i
)
Figure 10.
Accident Reactivity Shape Function, Cold Startup l
20 l
)
'5 23A5896 Rev. 1 1
20.0 17.5 1-15.0 i
s 12.5
-x 5
h 10.0
>-i
/
g 7.5
)
E i
5.0
)
2.5 A ACCIDENT l' UNCTION B BOUNDING UALUE280 CAL /G 0.0d.0 5.0 10.0 15.0 20.0 ROD POSIT 10N, FEET OUT
)-
l Figure 11.
Accident Reactivity Shape Function, Hot Standby 21
23A5896 Rev. 1 x
30.0 A SCRAM FUNCTION B B0VNDI WG VALUE 283 CAL /G F
25.0 9
ds
[
20.0
)
^
)
G 15.0 C
)
?
10.0 E35x
)
5.0
)
U 0 06'.o 1.0 fd 3.0 4.0 5.0 6.0
~'
ELAPSED TIME, SECONDS
);
)'
Figure 12.
Scram Reactivity Function, Cold Startup
).
23A5896 Rav. 1 50.0 A SCRAM FUNCTION h'
B BOUNDI NG VALUE 283 CAL /G 40.0 9
x h
30.0 8
G
)-
g i
20.0
~
5 C
)
W 10.0
[~
A.
).
0.06'.0 1.0 f.0 3.0 4.0 5.0 6.0 ELAPSED TIME, SECONDS I
).
Figure 13.
Scrain Reactivity Function, Hot Standby
)
23
)
23A5896 Rev 1 APPENDIX A L
LIMITING CONDITIONS FOR OPERATION This appendix provides the limiting condition for operation (LCO) for each of the power distribution limits identified below:
g (1) Average Planar Linear Heat Generation Rate (APLHCR)
(2) Operating Limit MCPR (3) APRM Setpoints
)
Surveillance requirements and required actions are specified in the Technical Specifications. The power distribution limit bases are given in Appendix B.
)-
A.1 APLHGR During steady-state power operation, the APLHGR for each type of fuel as
)
a function of axial location and average planar exposure shall not exceed limits based on applicable APLHGR limit values which have been approved for the respective fuel and lattice types determined by the approved methodology described in GESTAR-II (NEDE-24011-P-A). When hand calculations are
)
required, the APLHGR for_each type of fuel as a function of average planar exposure shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) shown in Figures A-1 through A-5, during two recirculation loop operation.
)
A.2 OPERATING LIMIT MCPR The applicable fuel cladding integrity safety limit MCPR for this cycle
).
is 1.04.
This safety limit MCPR applies to Unit 1 during this cycle because it is a D-Lattice BWR with at least two successive reloads of P8XBR, BPBXBR, CE8X8E, or GE8X8EB fuel types having high bundle R-Factors (>1.04), one of which is the fuel in its first cycle of operation. The use of this value has
)
24
)
23A5896-Rsv. 1
~been approved in Amencuent 14 of NEDE-24011-P A-8.
During steady-state power H
operation, the MCPR for each type of fuel shall not be less than the limiting.
value (shown.in Table A-1) times the Kg (shown in Figure A-6), for two:
recirculation loop operation.
In reference to Technical Specification 3.2.3.2, the 01)(CPR for r,y,
)
less than or equal to r3, is the greater of the non pressurization transient or the Option B OLMCPR (Table A-1), where
,,y, and,3 are given by:
a where:
1 - Surveillance test number.
)
n - Number of surveillance tests performed to date in the cycle (including BOC).
th
))
N -- Number of rods tested in the i surveillance test, g
- Average scram time to notch 36 for surveillance test i.
r g and
).
where:
)'
N - Number of rods tested at BOC.
y p - 0.813 seconds (mean value for statistical scram time distribution from de energization of scram pilot valve solenoid to pickup on y.
notch 36).
a - 0.018 seconds (standard deviation of the above statistical distribution).
)
25
)
I
)
23A5896 Rav. 1
.i, in reference to Technical Specification 3.2.3.2, the OLMCPR for r p
' greater'that r shall be either:
3 a.
The greater of the non-pressurization transient (Table'A-1) or the adjusted pressurization transient MCPR (MCPR adj ) * ***
- Option A Option B}
(
I MCPRadj ~'
Option B +
~
~
A7B
- 1.05. seconds (control rod average scram insertion rg time limit to notch 36),
and MCPR as given in Table A-1 Option A MCPR as given in Table A-1 0ption B or, b.
MCPR as given in Table A-1.
Option A A.3 APRM SETPOINTS
)
The-flow-biased APRM scram trip setpoint (S) and rod block trip setpoint (SRB) shall be:
S $ (0.66W + 54%) T, and S
$ (0.66W + 42%) T; RB where S and S are in per ent f rated thermal power; RB W - loop recirculation flow in percent of rated flow.
T is the. ratio of Fraction of Rated Thermal Power (FRTP) divided by Core Maximum Average Planar Linear Heat Generation Rate Ratio (CMAPRAT):
)
T
_FRTP _ where T $ 1.
CHAPRAT 26
)
fl
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jIIlt 1l
23A5896 Rev. 1 1.4
)-
)
1.3
)
1.2
)
AUTDMATic Flow cowTnot KI
)
1.'1 MANUAL PLOW CONTROL
)
scoor Tust stTPotNT i
CAutRATION P05mONED SUCH THAT FLOWMAX = 102 6%
107.0%
1.0 112.0 %
)
117.0 %
{
)
)
0.9 8
8 8
i i
1 30 40 60 60 70 80 90 100 CORE FLOW (%)
)
)
Figure A-6.
K Factor for CEXL-PLUS f
32
)
j
I
..f
'23A5896' Rev.'ll Table A 17 MCPRs:'
Fuel Type:
'P8X8R,,BP8X8R,' and GE8X8EB'
.Non-Pressurized Transient MCPR
-1.25-L Pressurization Transients
. Exposure Range-Option A Option B BOC7. to EOC7-2000 mwd /ST 1.32 1.25
)'
EOC7-2000 mwd /ST to EOC7 1.34 1,30 2
)J 2-
)f
);
)-
33
).
23A5896 Rsv. 1 APPEND 1X B R;
BASES FOR LIMITING CONDITIONS FOR OPERATION
.This. appendix provides the bases for each of the power distribution y'
limits identified in Appendix A.
B.1' APLEGR
)
This specification assures that the peak cladding temperature (FCT) following the postulated design basis loss-of-coolant accident (LOCA) will l
not exceed the limits specified in 10CFR50.46 and that the fuel mechanical design analysis limits specified in Reference B-1 will not be exceeded.
j
)
Thermal Mechanical Desien Analysis: NRC approved methods (specified in Reference B-1) are used to demonstrate that all fuel rods in a lattice oper-ating at the bounding power history meet the fuel design limits specified in
)
Reference B-1.. No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the. fuel design analysis APLHGR limit.
)
LOCA Analysis: A LOCA analysis is performed in accordance with 10CFR50, Appendix K to demonstrate that the permissible planar power (maximum APLEGR) limits comply with the ECCS limits specified in 10CFR50.46. The analysis is performed for the most limiting break size, break location, and single
)
failure combination for the plant. The methods used are discussed in Reference B-2.
The APLHGR limit is the most limiting composite af the fuel design
):
analysis APLHGR limit and the ECCS APLEGR limit.
B.2 OPERATING LIMIT MCPR
)
The required operating limit MCPRs at steady-state operating conditions as specified in Appendix A are derived from the established fuel cladding j
integrity safety limit MCPR specified in Appendix A and an analysis of j
i y
34 z____=_=-__________
)
i 33A5896 Rev. 1 l
j i
abnormal operational transients.
In the analysis of these abnormal opera-tional transients, the CEXL-PLUS thermal correlation has been used, where
)
4 applicable, to determine the appropriate initial conditions.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that
)
the resulting MCPR does not decrease below the safety limit MCPR at any time
)
during the transient, assuming instrument trip setting as given in Specifica-tion 2.2.1 of the Technical Specifications.
To assure that tre fuel cladding integrity safety limit is not exceeded
)
during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which ones result in the largest i
1 reduction in Critical Power Ratio (CPR). The type of transients evaluated were loss of fivv, increase in pressure and pever, positive reactivity
)
insertion, and coolant temperature decrease.
The codes used to perform the trrnsient analyses that serve as the basis for the operating limit MCPR are described in Reference B-1.
Conditions at
)
limiting exposures are used for nuclear data to provide conservatism relative to core exposure aspects.
Plant-unique initial conditions and system para-meters are used as inputs to the transient codes. The ACPR calculated by the transient codes is adjusted using NRC approved adjustment factors to account
)
for code uncertainties and to provide a 95/95 licensing basis.
The limiting transient yields the largest ACPR. The ACPR for the limiting transient is added to the fuel cladding integrity safety limit to
)
establish the minimum operating limit MCPR.
The purpose of the K factor is to define operating limits at other than f
rated flow conditions. At less than 100% flow, the required MCPR is the
)
factor.
Specifically, the K product of the operating limit MCPR and the Kg g
factor provides the required thermal margin to protect against a flow I
increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speedup caused by a motor generator speed controller failure.
35
)
23A5896' Rev. 1
~
factors assure For operation in the automatic flow control mode, the Kg that the' operating limit MCIR in Appendix A will not be violated should the most limiting transient occur at less than rated flow.
In the manual flow-control mode, the K factors assure that the safety limit MCPR will not be g
violated should the most limiting. transient occur at less than rated flow.
The K factor values are generically developed as described in g
References B-3 cnd B 4.
The K factors are conservative for the General Electric plant operation j.
g because the operating limit MCPRs in Appendix A are greater than the original l
1.20 operating limit MCPR used for the generic derivation of K.
g At core thermal power levels less than or equal to 25%, the reactor will g
be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the
) -
resulting MCPR value is in excess of requirements by a considerable margin.
With this low void content, any inadvertent core flow increase would only place' operation in a more conservative mode relative to MCPR.
During initial startup testing of the plant, a MCPR evaluation was made at 25% initial power
) -
level with minimum recirculation pump speed. The demonstrated MCPR margin was such, that future MCPR evaluations below this power level are unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient, since power distribution shifts are very slow when there have not been significant power or control rod changes. The y
requirement for calculating MCPR when a limiting control rod pattern is approached ensures that the MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.
j I
l 36
)
e-
33A5896 Rev. 1
- B' 3 APRM SETPOINTS The flow-biased thermal power upscale scram setting and flow-biased neutren flux upscale control rod birack functions of the APRM instruments are adjusted to ensure that fuel design and safety limits are not exceeded. The scram setting and rod block setting are adjusted in accordance with the
)
formula in Appendix A when the combination of Thermal Power and CMAPRAT in-dicate a highly peaked. power distribution. This adjustment may be accom-plished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM high flux scram curve by the reciprocal of the APRM gain change.
B.4 REFERENCES
)
1.
" General Electric Standard Application for Reactor Fuel",
NEDE-24011-P-A-8.
2.
" General Electric Company Analytical Model for Loss-of-Coolant Analysis y
in Accordance with 10CFR50 Appendix K", NEDO-20566, January 1970.
3.
Letter, J. S. Charnley (GE) to M. W. Hodges (NRC), " Application of i
GESTAR-II Amendment 15," March 22, 1988, MFN-027-88.
)
4.
Letter, A. C. Thadani (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Application of Amendment 15 to General Electric Licensing Topical Report NEDE-24011-P-A, ' General Electric Standard Application y
for Reactor Fuel' (TAC No. 60903)," May 5, 1988.
)
37 I
23A5896 RQv.'1 APPENDIX C p.
-PLANT PARAMETER DIFFERENCES GETAB and Transient Analysis Initial Conditions g
The values used in the CETAB and Transient Analysis which differ from p
the values reported in Tables S.2-4.1 and S.2-6 in NEDE-24011-P-A-8-US'are
'given in Table'C-1.
Table C-1 y
PLANT PARAMETER DIFFERNCES
),
Parameter Analysis Value NEDE-24011-P-A-8-US Value Rated Sterm Flow 10.47E+06 10.96E+06 0.2%
Dome Pressure 1005 1020 2 psi l
Turbine Pressure 950 960 2 psi Non-Fuel Power Fraction 0.039 0.040 Number of S/RVs 10 11
)
1 The indicated changes are a result of the appli::ation of the pre-approved methods outlined in Amendment 11 to NEDE 24011-P-A-8.
1
.The indicated change is a result of the simulation of a valve out-of-service condition.
38 1
23A5896' Rav. 1 APPENDIX D
-l 1
USE 0'F GEXL PLUS METh0DS FOR CYCLE 7 The analyses required for this cycle were performed with the GEXL-PLUS thermal correlation.
In analyses prior to Cycle 7 (Reload 6), the GEXL it
' thermal correlation was used. The incorporation of GEXL-PLUS into'the fuel cycle analysis' process is provided for in Amendment 15 to GESTAR-II
-(EEDE 24011-P-A-8). ~Any difference between this reload and the previous one l
'are due not only to cycle differences, but also to the difference in the j
)
methods. Therefore, making direct comparisons between the two cycles may be l
inconclusive.
).
21
)
-)
)
i
)
39
}
W 23A5896 Rev. 1 APPENDIX E j
it USE OF NEW SAFETY LIMIT MCPR FOR CYCLE 7 q
i The analyses required for this cycle were performed with the upgraded safety limit MCPR of 1.04, instead of the previous safety limit MCPR of 1.07.
y _-
The implementation of this safety limit is a result of the utilization of fuel types with high bundle R-factors, as stipulated in Ame.1dment 14 to
-GESTAR-II'(NEDE-24011 P-A-8).
Any difference between this reload and the previous one are due not only to cycle differences, but also to the
)
difference in the methods. Therefore, making direct comparisons between the two cycles may be inconclusive.
)-
)
)
)-
1 l
)
)
i 40
)
23A5896 Rev. 1 APPENDIX F I
MAIN STEAMLINE IS01ATION VALVE OUT-OF-SERVICE Reference F-1 provided a basis for operation of Brunswick Steam Electric Plant (BSEP) with one Main Steamline Isolation Valve Out-of-Service (MSIV00S)
(three steamline operation) and all S/RVs in service.
For this mode of operation in BSEP Unit 1 throughout Cycle 7, the EOC7-2000 mwd /ST to E007
'^
MCPR limits presented in Section 12 of the body of this report are bounding, and the peak steam line and peak vessel pressures-for the limiting Loverpressurization event (MSIV closure with flux scram) are 1228 psig and
)
1258 psig, respectively.
REFERENCE
)
F-1.
" Main Steamline Isolation Valve 0ut-of-Service for Brunswick Steam
~
Electric. Plant," EAS-117-0987, CE Nuclear Energy, April 1988.
)-
)
)
).
41
).
(Final) l
e.
p:~
l
i 1
)
. ENCLOSURE 5 l
i SUPPLEMENTAL RELOAD LICENSING REPORT BRUNSWICK SIEAM ELECTRIC PANT, UNIT 2 RELOAD 7, CYCLE 8 23A5855, REV. 1 t
l,-
I I