ML20247H851
| ML20247H851 | |
| 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 | |
| 23A5855, 23A5855-R01, 23A5855-R1, NUDOCS 8907310155 | |
| Download: ML20247H851 (45) | |
Text
.
23A5855 i Revision 1 Class I March 1989 f
i 1
);i (23A5855, Rev. 1)
SUPPLEMENTAL RELOAD LICENSING REPORT FOR
) BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 7, CYCLE 8
)
Prepared:
// Ag i> w LA 4 P. A. Lambeet
) Fuel Licensing Verified: L.
), . L. Rash
[FuelLicensing
) -
Approved:_J.S.Charnlef, Man 6ger Fuel Licensing GENuclearEneryr 3
175 Carm Annue SanJose. CA 95125 015 890725 r
$073l p DOC 05000324 "
D PNV
23A5855 Rev. 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY
)
This report was prepared by General Electric solely for Carolina Power and Light Company (CP&L) for CP&L's use with the U.S. Nuclear Regulatory Com-mission (USNRC) to amend CP&L's operating license of the Brunswick Steam ,
) Electric Plant 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 informa-tion in this document are contained in the Supplemental Agreement to the Con-tract between Carolina Power and Light Company and General Electric Company
) for Reload Fuel Supply and Related Services for Brunswick Steam Electric Plant Unit 2, and nothing contained in this document shall be construed as changing j said contract. The use of this information except as defined by said con-tract, or for any purpose other than that for which it is intended, is not
) authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any repre-sentation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or 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 from f such use of such information.
)
)
2 D
tb u
'Rev. 1 23A5855- .
F ACKNOWLEDGMENT
' The-engineering and reload: licensing analyses, which form the technical-basis of this Supplemental. Reload Licensing Report, were ~ performed by.T. P.
)J Lung and R. E. Polomik. of the Nuclear Fuel and Engineering Services Department.
p j.
?:
):
)
y
)-
3 Y
= _ - - - - - _ _ _ _ _ _ _ _ _ _
37 23A5855. Rev. 1 iU 1. PLANT-UNIQUE ITEMS (1.0)*
Appendix A:= Limiting Conditions for Operation-Appendix B: . Bases for Limiting Conditions for Operation 1- Appendix C: Safety Relief Valve Out-of-Service Appendix D:' Transient Operating Parameters Appendix E: Turbine Control Valve Configuration Appendix F: Use of GEMINI Methods for Cycle 8
)
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Type Cycle Loaded Number
)'
Irradiated P8DRB265H 5 44 BP8DRB299 6 184'
) BP8DRB299 7 148 New BD317A 8 92 BD323A 8 _9_2,
) Total 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
) Nominal previous cycle core average exposure at end of cycle: 20,449 mwd /MT Minimum previous cycle core nyerage exposure at end of cycle from cold shutdown considerations: 20,008 mwd /MT Assumed reload cycle core average exposure at end of
) cycle: 20,814 mwd /MT j 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 appropriate section in the United States Supplement, NEDE-24011-P-A-B-US, May 1986.
3 I
23A5855 Rev. 1 l l
l
) 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTE' - NO VOIDS , 20 *C (3.3.2.1.1 AND 3.3.2.1.2) l Beginning of Cycle, Keff
) Uncontrolled 1.114 Fully Controlled 0.968 Strongest Control Rod Out 0.988 l R, Maximum Increase in Cold Core Reactivity with 0.0
)
Exposure into Cycle, AK
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
)
Shutdown Margin ( AK) ppm (20*C, Xenon Free) 600 0.031
)
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(Cold Water Injection Events Only)
) 41.71 Void Fraction (%)
Average Fuel Temperature (*F) 1104 Void Coefficient N/A* (d/% Rg) -7.34/-9.18 Doppler Coefficient N/A* (4/*F) -0.205/-0.195
) **
Scram Worth N/A* ($)
)
- N = Nuclear Input Data, A = Used in Transient Analysis
) ** Generic exposure independent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986.
) 5
.-._.-__,_.____,._,.._.,__q 23A5855. Rev. 1~
( -7. ' RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS
-(S.2.2)
PeaMng Factors Fuel Bundle Power Bundle Flow . Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR I, Exposure: BOC8 to EOC8-2000 mwd /ST BP/P8x8R 1.20 1.55 1.40. 1.051 6.574- 111.5 1.22
'GE8x8EB 1.20 1.56 1.40 1.051 6.602 113.6 1.23
)
Exposure: EOC8-2000 mwd /ST to EOC8 BP/P8x8R 1.20 1.49 1.40 1.051 6.320 113.0 1.28 GE8x8EB 1.20 1.50 1.40 1.051 6.382 114.8 1.27 1
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization: No
) Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes Improved Scram Time: No
) Exposure Dependent Limits: Yes Exposure Points Analyzed: EOC8 and EOC8-2000 mwd /ST
- 9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3) j
) ~)
Single-Loop Operation: Yes 1
Load Line Limit: Yes Extended Load Line Limit: No
) Increased Core Flow: No Flow Point Analyzed: N/A !
- Feedwater Temperature Reduction: No ARTS Program: No j
> ' Maximum Extended Operating Domain: No f
)
23A5855 Rev. 1
! 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS' (S.2.2.1)
Methods Used: GEMINI-ACPR L Flux Q/A Transient (% NBR) (% NBR) BP/P8x8R GE8xBEB Figure Exposure Range: BOC8 to EOC8 Inadvertent HPCI 123- 119 0 15 0.15 2
)
Exposure Range: BOC8 to EOC8-2000 mwd /ST Load Rejection Without 379 118 0.15 0.15 3-Bypass
) Feedwater Controller 107 105 0.04 0.04 4 Failure Exposure Range: EOC8-2000 mwd /ST to EOC8 Losd Rejection Without 381 122 0.20 0.20 5
)' Bypass Feedwater Controller 153 107 0.05 0.05 6 Failure
) 11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT 3
SUMMARY
(S.2.2.1) !
Limiting Rod Pattern: Figure 7
)' Rod Block Rod Position ACPR MLHGR (kW/ft)
Reading (feet withdrawn) BP/P8x8R GE8x8EB BP/P8x8R GE8x8EB 104 4.5 0.15 0.15 14.08 15.08 105 5.0 0.16 0.16 14.52 15.52 ,
) 106 5.5 0.18 0.18 14.90 15.90 107 5.5 0.18 0.18 14.90 15.90 108 6.0 0.19 0.19 15.24 16.24 109 8.5 0.24 0.24 16.28 17.28
) 110 9.5 0.24 0.24 16.28 17.28 Setpoint Selected: 107
). 7
. E
23A5855 Rev. 1
- 12. CYCLE MCPR' VALUES (S.2.2)
Non-Pressurized Events BP/P8x8R GE8x8EB 1.
Exposure Range: BOC8 to EOC8 ;
inadvertent HPCI ' 1.22 1.22 Fuel Loading Error -- 1.20 Rod Withdrawal Error 1.25 1.25'
)
Pressurization Events
)
Option A Option B BP/P8x8R GE8x8EB BP/P8x8R GE8x8EB
' Exposure Range: BOC8 to EOC8-2000 mwd /ST
)-
Load Rejection Without Bypass 1,32 1.32 1.25 1.25 Feedwater Coptroller Failure 1.16 1.16 1.14 1.14
)_ Exposure Range: EOC8-2000 mwd /ST to EOC8 Load Rejection Without Bypass 1.33 1.33 1.29 1.29 Feedwater Controller Failure 1.17 1.17 1.14 1.14
)
- 13. OVERPRESSURIZATION A.NALYSIS
SUMMARY
(S.2.3)
P P,y y
) '
Transient (psig), (psig) Plant Response MSIV Closure 1213 1251 Figure 8 (Flux Scram)
)
8
).
'23A5855 Rev. 1 i 14. LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Disoriented Bundle Analysis Yes*'
)f- Event 6CPR Disoriented 0.11
- 15. . CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
).
. Bounding ' Analysis Results Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 I Scram Reactivity Functions: Figures 12 and 13 Plant-Specific' Analysis Results:
Resultant Peak Enthalpy, Cold: 139.4
)- Resultant' Peak Enthalpy, HSB: 192.7
- 16. STABILITY ANALYSIS RESULTS (S.2.4)
)- Rod Line Analyzed: Extrapolated Decay Ratio: Figure 14 Reactor Core Stability Decay Ratio, x2 /*0: 0.80 Channel Hydrodynamic Performance Decay Ratio, x2 !*0
) Channel Type BP/P8x8R 0.31 GE8x8EB 0.28
)' j
- 6CPR penalty of 0.02 for the tilted disoriented bundle is applied to the ;
cycle MCPR value reported in Section 12. !
)
9
):
1
i=
. -23A5855 Rev. 1-l
- 17. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2) !
)
See " Loss-of-Coolant Analysis Report for Brunswick Steam Electric Plant Unit No. 2," NEDO-24053, September 1977 (as amended), and NEDE-24053-P,
' August 1987.
1; 1., Fuel Type: BD317A (CE8x8EB)
Average MAPLHGR (kW/ft)
Planar y
Exposure Most Least Oxidation (CWd/St) Limiting Limiting PCT (*F) Fraction 0.0 11.23 11.65 2030 0.05 1.0 11.30 11.74 2065 0.06 2.0 11.49 11.87 -- --
)"
3.0 11.68 12.00 -- --
5.0 '12.07 12.38 2132 0.07 6.0 12.32 12.47 -- --
7.0 12.52 ~ 12.57 -- --
10.0 12.85 12.85 2194 0.08 15.0 12.65 12.66 2185 0.08
)
20.0 11.96 11.97 2092 0.06 35.0 10.03 10.05 1852 0.02 45.0 8.37 8.45 1691 0.01 50.0 5.91 5.99 1595 <0.01 y 2. Fuel Type: BD323A (CE8x8EB)
Average MAPLHGR (kW/f t)
Planar Exposure Most Least Oxidation (GWd/St) Limiting Limiting PCT (*F) Fraction
) 0.0 11.33 11.85 -- --
0.0 11.37 11.90 2013 0.05 1.0 11.54 12.05 2047 0.05 2.0 11.75 12.21 -- --
4.0 12.19 12.53 -- --
5.0 12.40 12.69 2173 0.08
) 6.0 12.62 12.69 -- --
7.0 12.68 12.72 -- --
15.0 12.72 12.73 2197 0.08 20.0 12.04 12.05 2121 0.06 25.0 11.38 11.40 2049 0.05 35.0 10.12 10.13 1868 0.03
) 45.0 8.49 8.55 1720 0.01 50.0 6.02 6.09 1605 <0.01 y 10
-J 23A5855 Rev.'l N
- ' .MMMMM.
- .MMMMMMMMM.
- MMMMMMMMMMM
- .MMMMMMMMMMM.
- MMMMMMMMMMMMM
':MMMMMMMMMMMMM r '::MMMMMMMMMMMMM
':E M M M M M M M M M M M M M l
':: M M M M M M M M M M M M E
': "MMMMMMMMMMM" l' MMMMMMMMMMM
> ; "MMMMMMMMM"
- "MMMMM" iIIIIIIIII
). 1 357 91113151'71921232527293133353739414345474951 FUEL TYPE A = BD323A D = BP8DRB299 (Cycle 6) )'
B = F8DRB265H E = BP8DRB299 (Cycle 7)
).
C = BD317A 1
)
Figure 1. Reference Core Loading Pattern 11
}-
23A5855 .Rev. 1 1 NEUTADN FLUX 1 VESSEL PRESS RISE (psi) 74 2 AVE SURFACE HEAT FLUX 4-44 2 RELIEF VALVE FLOW 4 3 CORE INLET FLOW 3 BYPASS VALVE FLOW 150 150 _
4 CORE INLET SUB 4 HPCI FLOW (% of tw)
)- 3 4
. g$ 2 h2 idk ,-2-1 .M 2 2 31 8.100 -3-3 : 3--3-3 3-i3 - 33 100 W
E
). - . g 50 50
(^4-4 4- 4 4-4-4 4-4 l--1 " 1 1
). 0 04 -te32-3 2 3-23 32 0 50 100 0 50 100 TIME (seconds) TIME (seconds)
)
1 LEVEL (inch REF-SEP SKRT) 1 VOlD REACTIVITY 2 VESSEL STEAMFLOW 2 DOPPLER REACTIVITY 150 .3 TURBINE STEAMFLOW -
i 3 SCRAM REACTIVITY 4 FEEDWATER FLOW 4 TOTAL REACTIVITY
) 5 h 1 1 1 1 100 c2-3-2-3 32-3 2 - 32 0 3 d 3"3-4--34-3-443-434 2 '~~~ ~ ~~~
4- 4 4 - 4 -4'4 ~4 '-4-4-4 T
- G p1 v--~-r :1 , 5 ~'
E 0 -2
). 0 50 100 0 50 100 TIME (seconds) TIME (seconds)
)
Figure 2. Plant Response to Inadvertent Startup of HPCI i
l
) 12
- m. . - .
3 23A5855 Rev. 1 I
~
1 NEUTRON FLUX 1 VESSEL PRESS RISE (psil 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FiOW I
3 CORE INi.,ET FLOV/ . 3 RELIEF VALVE FLOW 150 300 _,.
4 BYPASS VALVE FLOW
)
200 -j g : gpo 62' ) N2 .3 g [\ \ .1
- y'
'50 f) '2.
100 3
_3 kx 1
/ 3 I
O O 2 4 2M2- 4-2 4 4--4
) 'o 2 4 6 0 2 4 6
' TIME (seconds) TIME twoconds)
)
? LEVEL (inch-REF-SEPLSKRTI 2 VESSEL STEAMFLOW
/M 1
1 VOID REACTIVITY 2 DOPPLER REACTIVITY 200 3 TURBINE S'IAMFLOW .
3 8 3 SCRAM REACTIVITY 4 FEEDWATER FLOW 4 TOTAL REACTIVITY
- 4 2 e p 2 7
) % ., ./
) N 4 7 - 4 2-2 /
) ;- a y)
).. O
- 1 1
3 -3 3, ;3-3 g
l-1 \\
r
~
4 6 ~0 ii 2 4 6 O 2
)' TIME (seconds) TIME (seconds) 1:
Figure 3. Plant Response to Generator Load Rejection Without Bypass (EOC8-2000 mwd /ST) 1 13
I 23A5855 Rev. 1
- 1 NEUTRON FLUX 1 VESSEL PRESS RISE (psli 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 150 3 CORE INLET FLOW - 3 REllEF VALVE FLOW 4 CORE INLET SUB 4 BYPASS VALVE FLOW 4 100 L r j s \ y-4 .]
4
$ -4 1
-),v3-3k3---8 3 100. -3 3 4
{\2'\ \, . - -
I. . 50 f
b 01 6 342-34 3 2 -
1 0 '- 1 0 50 100 0 50 100 y
TIME (set ends) TIME (seconds) y 1 LEVEL (inch-REF SEP-SKRT) 1 VOID REACTIVITY 2 VESSEL STEAMFLOW 2 DOPPLER MEACTIVITY 3 TURBtNE STEAMFLOW _, 3 SCRASA REACTIVITY 150 .
- ,, 4 TOTAL REACTIVITY 4 4- 4 FEEDWATER FLOW
> g J
! , /
? 2 ( 1-100 O' 2 ""---- 2 M " 2
. 2 & i 1 5 V '
4 50 -1 e
/ :
\'"
4
\
g O 3-3-3b.3 100
-2 0 50 100 0 50 l TIME (seconds) TIME (seconds) l Figure 4. Plant Response to Feedwater Controller Failure (E0C8-2000 mwd /ST)
) 14 L
l
_ -= _____-_____
- - .1-23A5855 Rev. 1
)
2$ht $$r'AE!'m At rLux h hN!v $O!!$I "
L*
300
'150 L
I 100
, d b-y
-- 200 e N 60
' 100- A C - ; ;
0 k~ ' '
O - : : : : .
4 6 0 2 4 6
) 0 2 TIME (seconds)
TIME taeconds)
)
/\ / 1 V01D REACTIvliv 1LEVEttINCEREF-SEP.$nRT) 2 DDPFLtR RI A:TIV11f 2 VESSEL $TEAWLDv 200
" ~ 1 y
) .,
N
> \ \
100- . - , , 0 -
7
/
'M~ - -
u j
! \ R
)- -,
W 5 0 {fy :
- ; ; ; -1 - - - -
Y
/
)J -100 -2 -
4 6 l 4 6 0 2 0 2 TIME (seconds)
TIME (seconds)
{
l
)
Figure 5. Plant Response to Generator Load Rejection Without Bypass (EOC8)
) 15 J
i -
1 n i I
-23A5855 Rev. I 1
)c 1 NEurRDN FLUX l VEE:EL PRf55 RISE (PSI) 2 ave SURFAct of AT FLUX 2 SAF :17 VALVE FLCa 3 EDR' 1NLET FLOW 3 REL lEF WALVE FLOW
' " "' '"& 4 erei.ss VALVE FLO 150 t b '
100 '-
_ME AY 8 100 , - -
.: ;7" Y -
W 50
)..
so
( 3 OM.d : ;; = c ;^ :: :
p_ - __
)' O 10 20 0 to 20 TIME (seconds) TIME (seconds)
)
' ! Vol) REAttitITYI i1Ev :L ( INC64.REF.$E P.$nRT ) g 2 VES EL STEAMFLDw 2 00P'LER RE All1)V!1Y gpo
?$$'I"730" g ! !!*]Ef*E !!!
) G V
m y
' ' ^^ ^
gg, -
E OL- -- -T. --
'.fu 50 #
/ !-1 sc i
0 : ; ; -2
): 10 0 10 20 O 20 TIME (seconds)
TIME (seconds)
)
Figure 6. Plant Response to Feedwater Controller Failure (EOC8)
I
) 16
23A5855 Rev. 1 i
).
2 6 10 14 18 22 26 30 34 38 42 46 50
) 18 18 52 48 36 44 44 44 36 44 14 2 10 10 2 14
}
40 44 44 44 44 44 2 10 2 2 10 2 36 44 44 44 44 32
) 10 2 0 0 2 10 28 44 44 44 44 24 20 2 10 2 2 10 2
) 44 44 44 44 44 16 14 2 10 10 2 14 12 8 36 44 44 44 36
) 4 18 18 NOTES: 1. NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.
) 2. ERRORR0 DIS (22,28).
)
)
Figure 7. Limiting Rod Pattern
, u
l 23A5855 Rev. 1
)
)$dEUTRON F UX i VESSEL PU55 R!$E(PSI) 2 AVE $URF A:t N AT FLUX 2 $AFETY WA.vE FLCv 3 CORE ANLET FLOW 3' RELIEF VA.VE FLOW 300 9""'** "E 'L9" 150 I. -
i$ 100 .~~ 200
^
50 100 f 2 ;
')
).. O \- 0;: J
_ :n _ ; :
6 0 6 0 TIME (seconds) TIME (seconds)
)
eLEVEtt!NC4. RET.EEP.SKS T ) 1V D REACTIVITY 3 h8$NE E W 'Y' dR E 200 '
- E E ^- "r " 'L 5 1 * ~ ^L at--" "-
_ V
) :
s2 2
100
-A
, p 0..-
'N 8 u
) 5 -1 D 3
0 : : : @ -1 6
u:
-100 -2
)
TIME (seconds) TIME beconds) 1
)
Figure 8. Plant Response to MSIV Closure - Flux Scram i
) 18
i' ~
23A5855 Rev. 1' .
i
), "
- 0. 0 1
1 -5. 0
'l l
-10.0 Jb i o -15.0 i
)
s-- t b- .
C -20.0 / --
E
).$
8 -25.0' __
)
s
/
(
o U
-30.0
) -
A CALLL LAlt.U VALUE -COLD BCALCL LATED VALUE -HSB C BOUNE VAL 280 C/ L/G COLD D BOUNE VAL 280 C/ L/G HSB jl -40.0 ,_
- 0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.
?.
Figure 9. Fuel Doppler Coefficient in 1/6'C
) 19
l 23A5855 Rev. 1 ;
1 20.0 l J
l i
i 17.S i
15.0
) m O
3 12.5 ;
i r '
)
- _a t,
g 10.0 -
l, t, ;
s" w
)
> 7.5 t-o w
5.0
)
2.5 A ACCIDENT FUNCTION l
) B BOUNDING '/ALUE 280 CAL /G
- 0. 0 ,;-
- 0. 0 5.0 10.0 15.0 20.0 ROD POSITION, FEET DUT j
) l
)
Figure 10. Accident Reactivity Shape Function - Cold Startup
) 20
23A5855 Rev. 1 1 1
20.O i
)-
17.5 15.0 y-S, ,,
W 12.5 o 7 >
)
_J g 10.0
) C > 7.5 .
b w
Q"
) 5.0 g 2.5
)
A ACCIDENT FUNCT]ON B BOUND]NG 'fALUE 280 CAL /G
- 0. 0 f
- 0. 0 5.0 10,0 15.0 20.0
)
ROD POSITION, FEET DUT
)
Figure 11. Accident Reactivity Shape Functica - HSB 21
23A5855 Rev. 1 30.0 A SCRAM FLNCilON B BOUNDING VALUE 280 CAL /G
)
25.0
)
c w "
- 20.0 r
-H
) Ed a
n 15.O o /
!$$ /
v
) >
b /
> 10.0 s
O
') <
w a:
5.0
)
- 0. 0 , v:, d
- 0. 0 1. 0 2.0 3.0 4.0 5. 0 6.0
)
ELAPSED TIME, SECONDS
)
Figure 12. Scram Reactivity Function - Cold Startup
) 22
23A5855 Rev. 1
o
) I w
x
< 30.0 t--
)' W a
m a
w z
v
) o 20.0 t-
> s
.-i o
) w cc 10.0
/
) l t
1 s
0, 0 ,, j---
^M
- 0. 0 1. 0 2.0 3. 0 4.0 5.0 6.0
)
ELAPSEU TIME, SECONDS .
)
Figure 13. Scram Reactivity Function - HSB I I
) 23
7
.A:
23A5855 Rev. 1
^
- L
1.25 A NATURAL CMCULATION B 105 PERCENT ROD UNE 1,00
)-
,A' O.75 ,
o AB
- w. /
). .
1 8
s 0.50 )
t, 0.25
)
i_B C .
0 20 40 SO 80 100 120 3 POWER (%)
)
Figure 14. Reactor Core Decay Ratio Versus Power 24
)~
23A5855 Rev. 1 APPENDIX A LIMITING CONDITIONS FOR OPERATION This appendix provides the limiting condition for operation (LCO)~ for
)
each of the power distribution limits identified below:
(1) Average Planar Linear Heat Generation Rate (APLHGR)
(2) Operating Limit MCPR
} (3) APRM Setpoints Surveillance requirements and required actions a re specified in the Tech-nical Specifications. The power distribution limit bases are given in Appen-I dix B.
A.1 APLHGR
)I 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-4, during two recirculation
) loop operation.
A.2 OPERATING LIMIT MCPR
) The fuel cladding integrity safety limit MCPR is 1.07. During steady-state power operation, the MCPR for each type of fuel shall not be less than the limiting value (shown in Table A-1) timer the Kg (shown in Figure A-5),
for two recirculation loop operation. l
) ,
kh
gg , ' "
23A5855 Rev.'1 l i
i i! -In reference to Technical Specification 3.2.3.2, the OLMCPR for T,y, less -
than or equal to T is the greater of the non pressurization transient or B
the Option B OLMCPR (Table A-1), where T,y, and TB *#* 8 **" Y*
i
)1 n'.
-[ N g Tg 7 .. i=1 ave n
.[i=1N'g
)!
where:- !
~ i = Surveillance test number. I i
n = Number of surveillance tests performed to date in the cycle (includ-ing BOC)..
f Ng = Number of rods tested in the i ' surveillance test.
'l Tg = Average scram time to. notch 36 for surveillance test i.-
); and I TB " M + 1.65 ,
[N, g
o i1 1
where:
y Ny = Number of rods tested at BOC.
j
^'
11= 0.813 seconds (mean value for statistical scram time distribution l i
from de-energization of scram pilot valve solenoid to pickup on i l
y i notch 36). ]
i l
o =. 0.018 seconds (standard deviation of the above statistical {
distribution).
26
_ _ _ _ _ _ - - _ _ - 1
- t :
e 23A5855 Rev. 1
- t. In reference to Technical Specification 3.2.3.2, the OLMCPR for T greater than T shall be either:
B
- a. The greater of the non-pressurization transient (Table A-1) or the .
). adjusted pressurization transient MCPR (MCPRadj).where:
T T
= PR ~
adj 0ption B + T -T B
ption A Option B TA = 1.05 seconds (control rod average scram insertion l time limit to notch 36).
and
)
MCPR O @ A as given in Table A-1 MCPR as given in Table A-1 0ption B
)- or,
- b. MCPR as given in Table A-1.
OWuA
) A.3 APRM SETPOINTS The flow-biased APRM scram trip setpoint (S) and rod block trip setpoint (SRB) shall be:
S 1 (0.66W + 54%)T, and E
RB 1 (0.66W + 42%) T;
) where S'and SRB are in percent of rated thermal power; W E loop recirculation flow in percent of cated flow.
T is the ratio of Fraction of Rated Thermal Power (FRTP) divided by Core Maxi- ,
num Average Planar Linear Heat Generation Rate Ratio (CMAPRAT): i T=C T where T 1 1.
)- 27 i
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) Figure A-5. K Factor f
) 32 l
i
n--_-. . . .
- , i /;
23A5855' Rev. 1 A
- j. Tame A-l' MCPRs F
Fuel Type: P8x8R, BP8x8R, and GE8x8EB p
Non-Pressurized Transient MCPR = 1.25 Pressurization Transients
)I.
Exposure Range. MCPkptionA MCPbptionB BOC8 to EOC8-2000 mwd /ST 1.32 1.25'
) .-
EOC8-2000 mwd /ST to EOC8 1.33 1.29 1
).
i y
). ,
f 7
) 33 l
1
23A5855 Rev. 1
)
APPENDIX B BASES FOR LIMITING CONDITIONS FOR OPERATION This appendix provides the bases for each of the power distribution
) limits identified in Appendix A.
B.1 APLHGR i
) This specification assures that the peak cladding temperature (PCT) f ol-lowing the postulated design basis loss-of-coolant accident (LOCA) will not !
exceed the limits specified in 10CFR50.46 and that ^,he fuel mechanical design analysis limits specified in Reference B-1 will not be exceeded.
)
Thermal Mechanical Design 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 APLHGR) limits comply with the ECCS limits specified in 10CFR50.46. The analysis is performe d 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 APLEGR limit is the most limiting composite of the fuel design analy-sis APLHGR limit and the ECCS APL2iGR limit.
f B.2 OPERATING LIMIT MCPR The required operating limit MCPRs at steady state operating conditions j
) as specified in Appendix A are derived from the established fuel cladding integrity safety limit MCPR specified in Appendix A and an analysis of
) 34 l
23A5855 Rev. 1 i
abnormal operational transients. For any abnormal operating transient analy-sis 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, assum-
) ing instrument trip setting as given in Specification 2.2.1 of the Technical Specifications.
To assure that the fuel cladding integrity safety limit is not exceeded
) during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which ones result in the largest reduc-tion in Critical Power Ratio (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion,
) and coolant temperature decrease.
The codes used to perform the transient 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 param-eters are used as inputs to the transient codes. The ACPR calculated by the j transient codes is adjusted using NRC approved adjustment factors to account f 1
) for code uncertainties and to provide a 95/95 licensing basis.
The limiting transient yields the largest 6CPR. The 6CPR for ;.he limiting transient is added to the fuel cladding integrity safety limit to
) MCPR to establish the minimum operating limit MCPR.
l I
The purpose of the gK f actor is to define operating limits at other '
than rated flow conditions. At less than 100% flow, the required MCPR is the
) product of the operating limit MCPR and the Kf factor. Specifically, the Kg factor provides the required thermal margin to protect against a flow ]
increase transient. The most limiting transient initiated from less than j rated flow conditions is the recirculation pump speedup caused by a motor
) generator speed control failure.
l l
) 35
i 23A5855 Rev. 1
)
For operation in the automatic flow control mode, the K factors assure I f
thatlthe operating limit MCPR 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 f
I violated should the most limiting transient occur at less than rated flow. l The Kf factor values are generically developed as described in Reference B-1.
). -
The K f actors are conservative for the General Electric plant opera--
f tion because the operating limit MCPRs in Appendix A are greater than the original'1.20 operating limit MCPR used for the generic derivation of Kf .
) .:
At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum. recirculation pump speed and the moderator void con- j tent will be very small. For all designated control rod patterns which may be
) employed at this point, operating plant experience indicated that the result-ing 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 l
) startup testing of' the plant, a MCPR evaluation will be made at 25% initial power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above
) 25% rated thermal power is sufficienc, since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or 1 power shape, regardless of magnitude that could place operation at a thermal limit.
)'
) 36
23A5855 Rev. 1 B.3 APRM SETPOINTS The flow-biased thermal power upscale scram setting and flow-biased neu-tron flux upscale control rod block 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 for-mula in Appendix A when the combination of Thermal Power and CMAPRAT indicate a highly peaked power distribution. This adjustment may be accomplished by increaging 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 (latest approved revision).
)
- 2. " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K", NED0-20566, January 1970.
)
)
)
)
) 37
i 23A5855 Rev. 1
) APPENDIX C SAFETY-RELIEF VALVE OUT-OF-SERVICE The analysis was performed for safety-relief valve out-of-bervice and I there was no change in the 6CPR. The change in pressure for MSIV with flux scram is shown below.
P,7 P y
) (psig) (psig) Plant Easponse MS1V Closure 1225 1261 Figure C-1 (Flux Scram)
)
)
)
)
)
)
) 38
o
)
23A5855 Rev. 1 1
1 NEUTRON FLUX 1 VESSEL P'RESS RISE (psi) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW
{ 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 35 0. 0 300.0 4 BYPASS VALVE FLOW 2Y 300.0 2- Y 8 3_2 200.0 g .2 b ' %'N' 3
3 50.0 2 300.0 --
3-3 3 3- -- 3
- 8*8 i
% 43 324242 24--4-2-4 4
- 0. 0 5.0 0.0 5.0 TIME (SECONDS) Tit'E (SECONDS) 1 LEVEL (inch-REF-SEP SKRT) \ 1 VOID REACTIVITY 2 VESSEL STEAMFLOW 2 DOPPLER REACTIVITY 2 U. 4 3 TURBINE STEAMFLOW -
s.0 3 SCRAM REACTIVITY 4 FEEDWATER FLOW / 4 TOTAL REACTIVITY h 4 H. 3. 0 , : 4 4 A .4-4 4 s.0 y, _43 ,
N4 /
2 3'3 1 2h 42 4 %2 k b 2 3 .
3 C 80 3x
}2 3 3
=
.s.O 3
4 ,
)
300 0 -2.0 0.0 5. 0 0. 0 5.0 TIME (SECOND$1 TIME (SECONDS)
)
Figure C-1. Plant Response to MSIV Closure - Flux Scram (SRV005) 2.
39
I i.
23A5855 Rev. 1 1
1 APPENDIX D. I PLANT PARAMETER DIFFERENCES GETAB and Transient Analysis Initial Conditions
)-
The values used in the GETAB and Transient Analysis are given in Table D-1. The following values differ from the values reported in Tables S.2-4.1'and S.2-6 in NEDE-24011-P-A-8-US, May 1986.
I
)
Table D-1 PLANT PARAMETER
)
Parameter Analysis Value NEDE-24011 value Dome Pressure 1005 1020 1 2 psi
) Rated Steam Flow 10.47 10.96 + 0.2%
Turbine Pressure 950 960 t'2 psi
. Non-Fuel Power Fraction 0.039 0.04
)'
)
}l
)
) 40
23A5855 Rev. 1
(
) APPENDIX E TURBINE CONTROL VALVE CONFIGURATION The transient GETAB analyses presented in the body of this report are
);; based on turbine control valves in a full-arc configuration and on the power supply to the recirculation Motor-Generator Sets from offsite power.
)
). -
)
)
)
)-
41 7'
)
23A5855 Rev. 1 i
,. APPENDIX F USE OF 3 31NI METHODS FOR CYCLE 8 .
1 The analyses required for this cycle were performed with GE's advanced-
)
reload licensing methods, known as GEMINI. Any differences 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 will be inconclusive.
)-
-i
)
i 1
)
)
1 42
) (FINAL) i
_ _ __u______ _ ____--___ ._z---------__---__--------_-----_---------- - _ _ - - - _ - - - - -- - - - - - - - - _ - - - _- - ---- - J