ML20113H214
| ML20113H214 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/30/1984 |
| From: | Charnley J, Dennison D, Zarbis W GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20113H173 | List: |
| References | |
| 23A1806, 23A1806-R, 23A1806-R00, NUDOCS 8501250041 | |
| Download: ML20113H214 (28) | |
Text
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23A1606 NOVEMBER 1984 i
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) I SUPPLEMENTAL RELOAD LICENSING i SUBMITTAL FOR
- JAMES A. FITZPATRICK -
j NUCLEAR POWER PLANT RELOAD 6
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23A1806 Revision 0 Class 1 November 1984 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT RELOAD 6 Prepared: //k1 - %
- 6. K. DennTson Verified: M W. A. Z r is
//
Approve : L ef / E /d e g V
NUCLEAR ENERGY BUSINESS OPERATIONS
- GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 9512$
GENER AL $ ELECTRIC 1/2
'1A1806 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY
' This' report;was pcepared by General Electric solely for The Power Authority of the State of New York (The Authority) for The Authority's use with the U;S. Nuclear Regulatory Commission (USNRC) for amending The Authority's
- operating license of the James A. FitzPatrick Nuclear Power Plant. 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 The Authority and General Electric Company for nuclear fuel and related services for the
. nuclear system for The James A..FitzPatrick Nuclear Power Plant, dated August 1, 1981, 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, Lis'not 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 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 such use of such information.
1 3/4 r ,
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23A1806 Rav. 0
- 1. ' PLANT UNIQUE ITEM (1.0)*
Appendix A: CETAB Analysis Initial Conditions 2.: RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel' Type Cycle Loaded Number Number Drilled Irradiated P8DRB284H 5 112 112 P8DRB299 5' 52 52 P8DRB299 6 200 200 New BP8DRB299 7 196 196 Total- 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of' cycle: 17821 mwd /ST Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 17621 mwd /ST Assumed reload cycle core average exposure at end of cycle: 18178 mwd /ST.
Core loading. pattern: Figure 1
'*(=) Refers.to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-6. A letter "S" preceding the number refers to the appropriate country-specific supplement.
5
' #- 9 r l
).5 : i f ,? $
' 23A1806 Rav. O' 14.3 . CALCULATED' CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH ~- NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)
Beginning of Cycle K,gg
- Uncontrolled 1.116 Fully Controlled
~
0.961 Strongest-Control Rod Out 0.989
- R, Maximum Increase in Cold Core Reactivity 0.001
-with Exposure into Cycle, AK
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3).
/
Shutdown Margin (Ak)
. jggs - (20*C, Xenon Free) 600 0.026 6.J RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2) ,
(REDY EVENTS ONLY)
EOC-1000 'EOC-2000
'EOC mwd /ST- mwd /ST Void' Fraction (%) 41.7- 41.7- 41.7 M ' Average Fuel- 1270. 1270 1270 Temperature ('F)-
, Void Coefficient -9.06/-11.33 -9.80/-12.25 -10.25/-12.82
,N/A* (c/% Rg)
~ Doppler Coefficient -0.230/-0.219 -0.227/-0.216 -0.222/-0.211-
- :N/A (c/*F):
' Scram Worth N/A ($)
- n. , ^
- N =, Nuclear: Input Data A'= Used in~ Transient Analysis
- Generic' exposure. independent. values are used as given in " General Electric Standard Applicatioi for Reactor Fuel," NEDE-24011-P-A-6, April 1983.
6
~ . _ _ - - . . _ _ _ . _ . ' _ _ - . _ . - - . - - - - - - _ - . - - - _ - . - . . - - - - , _ - _ - . - _ _ _ . - - _ - - - - - - _ _ _ _ . _ _ - - ' '
23A1806 Rsv. 0
- 7.; RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Fuel ' Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposure: EOC '
B/P8x8R l 1. 20 ' 1.41 1.40 1.051 6.004 117.2 1.35 Exposure: _EOC-1000 mwd /ST B/P8x8R 1.20- 1.44 1.40 1.051 6.144 116.4 1.32 Exposure: EOC-2000 mwd /ST B/P8x8R 1.20 -1.48 1.40 1.051 6.291 115.5 1.29
- 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: Yes (ODYN Option B)
- 9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)
-Single-Loop Operation: Yes Load Line Limit: Yes
. Extended Load Line Limit: No
. Increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No 7
n.
23A1806 Rsv. 0
- 10. L)RE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
. Exposure Flux Q/A ACPR Transient (mwd /ST) (% NBR) (% NBR) B/P8x8R Figure Load ' Rej ection EOC 624 129 0.28 2a Without Bypass EOC-1000 535 126 0.25 2b E0C-2000 450 124 0.22 2c Loss of 80'F E0C 124 121 0.15 3 Feedwater Heating Feedwater Controller EOC 469 128 0.25 4a
. Failure E0C-1000 390 126 0.23 4b EOC-2000 312 121 0.18 4c
- 11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(S.2.2.1)
Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: Yes Rod Block Rod Position ACPR MLHGR (kW/ft)
Reading '(ft withdrawn) B/P8x8R B/P8x8R 104 3.0 0.15 16.78 105 3.5 0.17 17.50-106 4.0 0.20 17.93
-107 4.0 0.20 17.93 108 4.5 0.22 17.96 109 5.0 0.23 17.96 110 6.5 0.27 17.96 Setpoint Selected: 108 8
23A1806 Rzv. 0 12.. CYCLE MCPR VALUES (S.2.2)
B/P8x8R Non-Pressurization Events Exposure-Range: BOC to EOC
. Loss of 80*F Feedwater Heating 1.22 Fuel Loading Error 1.20 Rod Withdrawal Error 1.29 Option A Option B B/P8x8R B/P8x8R Pressurization Events Exposure Range:
'BOC to EOC-2000 mwd /ST-Load Rejection Without Bypass 1.35 1.14 Feedwater Controller Failure 1.31 1.24
. Exposure Range:
EOC-2000 mwd /ST to EOC-1000 mwd /ST Load Rejection Without Bypass 1.38 1.16 Feedwater Controller Failure 1.36 1.29 Exposure Range:
E0C-1000 mwd /ST to EOC Load Rejection Without Bypass 1.41 1.29 Feedwater Controller Failure 1.38 1.31
- 13. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3) sl v Transient (psig) (psig) Plant Response MSIV Closure 1218 1255 Figure 6
((Flux Scram) 9
n 23A1806 Rev. 0
- 14. STABILITY ANALYSIS RESULT (S.2.4)
- Rod Line Analyzed: Extrapolated Rod Block Decay Ratio: Figure 7
. Reactor Core Stability Decay Ratio, x2 /*0 0.86
' Channel' Hydrodynamic Performance Decay Ratio, x2 /*0
~
Channel Type B/P8x8R' O.30 e
,, 15. LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes
- Event Initial CPR Resulting CPR
- Misoriented 1.18 1.07 16; ~ CONTROL-ROD DROP ANALYSIS RESULTS (S.2.5.1)
- Bounding Analysis Results:
Doppler Reactivity Coefficient: Figure 8
- Accident Reactivity Shape Functions: Figures 9 and 10
~
-Scram Reactivity Functions: Figures 11 and 12 Plant Specific Analysis Results:
Parameter (s) not Bounded, Cold: All parameters' bounded-Resultant Peak Enthalpy, Cold: .N/A Parameter (s).not Bounded,.HSB: All parameters baunded Resultant Peak Enthalpy, HSB: N/A
'17. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2) g- See " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)," July- 1977, NEDO-21662 (as amended) .
10 L
__ ._1____m. _ _ . . . - . . _ _ _ _ _ _ _ _ _ . _ _ . ___.__._______.____.m __
23A1806 Rev. 0
- sMBIBIBsBis
- :" .BEBBBsBBBEMBsBEBis i :: BEBsB8BIBEMMBIBBBEM
- :: eBsBEMBBBBBBBBBBBBBBBis
- BsBEBEBEBBBEMBEBBBBBsBEBs i
- BsBE988E88B888BBBEBsBBBsBE
- "
- MBEBEMBsBBBEBEMBEBaBsBE
! ':BsBEBEBEBBBsi+2BsBsBsBsBEBE
- ::MBBBsBBBBBBBBBBBBBBBBBsBE
- *BEBBBEMBBBBBBBBBBBEBi"
- :' B8BsBEBsBBBBBBBBBEBsBE "MMMMs+2NMME"
- "MMMMM" IIIIIIl l l l c 1 357 9111315171921232527293133353739414345474951 FUEL TYPE a
i A = P8DRB284H B = P8DRB299
{
C = P8DRB299
- D = BP8DRB299 U
, Figure 1. Reference Core 1~ading Pattern 11 b
t 23A1806 Rev. 0 I NEUTRON FLut i VESSEL PRESS RISE (PSI) 2 AVE SURFACE KAT FLUX 2 SAFETY VALVE FLOW 3 CORE IttET TLOW 3 RELIEF WALVE FLOW 150,0 300.0 eevonss u_ituE rLnu
[
j 200.0 A %
N 200.0
~
n \ %
- 50.0 l N 100.0 -
' O. 0 __ . . 0.0 , , , , , ,, , ,, , , ,
~... - . . . . -
6.0 0.0 2.0 4.0 6. 0 ;
0.0 2.0 4.0 TIME (SECONOSJ TIME (SECONOS) 1 LEVEL (INCH-REF-SEP-SKRT) 1 10 REACTIVITV 2 VESSEL STEAFLOW 2 PLER REACTIVITY 3 TURBINE STEnWLOW 3, SCR REACTIVITY e_ n_ , 1.0 v. n_ r i orir_r,urr.y 200.0 ' e e_ r_ n_ u i v. r_ o. .
U m- - a A 5 88 - I /
100.0 h b v . .
. A
.~
. - . g
- 0. 0 ,g; , , , , , , h .l.0 g
g -v -
.i00.0 2.0 1 2.0 4.0 8.0 0.0 2.0 4. 0 8.0 0.0 fine (SECONOS) TIPE (SECONOS)
Figure 2a. Plant Response to a Load Rejection Without Bypass (EOC) 12
i.
23A1806 R:v. O b 1 NEUTRON FLU ( l VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOV 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 150.0 300.0 ' eyn.*SS uAtuE rLey 100.0 l 200.0 m: -\
b '
N E k
" 50.0 4
% ago,e 4-3 0-- 0 0 ~
- 0. 0 _ . 0.0 , , , , , , ,, , , , , ,
- 0. 0 2. 0 4.0 6. 0 0.0 2. 0 4. 0 6. 0 TIME (SEC080S1 TIME (SECONOS)
I LEVEL (INCH-REF-SEP-SKRT) 1V D REACTIVITY 2 VESSEL STEA9 FLOW 2 LER REACTIVITY 200.0 ' E M "L S'! "f'
- i.0 v
' S'" ^
--- - "55!!!!!! -
N f
100.0 'W - a A 0.0
. _ . /
oN3/v yrt% g t N '
+[,)
F_
- 0. 0 ,: , , , , ,,, y 3,g
- v * * - - - - -
g
\
-100.0 -2.0 C. 3 2.0 4.0 6. 0 0. 0 2.0 4.0 6.0 TIME (SECONOS) TIME ($[CONOSI Figure 2b. Plant Response to Load Rejection Without Bypas (E0C-1000 mwd /ST) 13
23A1806 Rev. 0 U 1 NEUTRON FLU ( 1 VESSEL PRESS RISE (PSI) 2 AVE SURF ACE HEAT FLUX 2 SAFETY V ALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 150.0 300.0 '. nyons? ua'_rE etcy
^
100.0 ; - .- I 200.0 a ~
7 b N z :
50.0
\ N 100.0
, _e I
- 0. 0 . , 0.0 _ , , , , , ,, ,, , _
c
- 0. 0 2.0 4.0 6.0 0.0 2. 0 4. 0 6.0 TIME (SECONOS) TIME (SECONOS)
I i
1 LEVELt!NCH-4EF.SEP.SKRT) / 1V O REACilVITY ER REACTIVITY 2 VESSEL STEA* FLOW T 2 l
3 TURBINE STEAMFLOW 3 SCRA REACTIVITY E 1.0 m vnv ai. crari r. v e r. v.
200.0 m r r en_ u n Tr_ o. e--.nu
-l r
O
- - ^ 5 88 100.0 *-
W l
NV vbe vi ,
i8 1 i 'P %
g7 g O l
- 0. 0 , , , , , _ _ $d 1. 0 1 5 , . q . . . . . . -
g ;
L f 300.0 2.0 I 2.0 4.0 6.0 0.0 2. 0 4. 0 6.0 O. 0 TIME (SECONOS) TIME (SECONOS) s r
=
Figure 2c. Plant Response to Load Rejection w/o Bypass (EOC-2000 mwd /ST)
F 14 k
i
23A1806 Rev. 0 150.0 NE RON UX 1 VESSEL PRESS RISE (PSI) 2 AVE S CE HEAT FLUX 2 RELIEF VALVE FLOW 3 COR"1 FLOW 3 BYPLSS VALVE FLOW 130.0 ' e n=
' I u_ _ ?T l
100.0 8
W' I
$ .10 0. 0
.I b
g 50.0 u
li!
t-00.0 ,
C. 0 l 0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECONOS) TIME (SECONOS)
I LEV ELCINCH.REF.SEP.SKRT) i VOI ) REACTIv!TY 2 VES :EL STEANFLOW 2 00F > ER REACTIVITY 139.0 3 TUR31NE STEAMFLOW 1.0 3 SCE, n REACTIVITY
' rEE wATE PLOu i eg n _ ngitt3v3vv A h
- z 0.0 IH.e ;. 3; _
- . 7 e
8 G
!:. t ,
ti! . . 0 e.: ,L, d..
2.0
- f. 0 103.0 200.0 0. 0 133.0 200.0 f!ME (SECONO3) TIME (SECONOS)
Figure 3. Plant Response to Loss of 80*F Feedwater Heating 15
__ j
g n p l3m l p u,
- r .ag ;p l i v-.l u
- l I' rlm qq[." n"4PERR N i.E' I i p RR R I 4 4R p' 23A1806 Rev. O 150.0 1 NEurRON FLUX 1 VES 3EL PRESS HI ( $!)
2 AVE SURFACE H! (T FLUX 2 Stf ETY YALVE FL 3 CORE INLET FL] i 3 REL [EF VALVE FL1 W 4 BYP (SS VALVE FL)W
~
e cae r '" et c' 150.0
_ l 100.0 L /k ,
pi '
Q 8 -- K :- : -. -
l I" 10 0. 0 < ; "' ; O 0 N
s
^
w 50.0
- 5 M
l t I
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n 50.0
= . J 0.0
- g M4I; C 0^ 'O 32 ]'~ '
i I 0. 0
! O. 0 10.0 20.0 0. 0 10.0 20.0 TIME (SECONDS) TIME (SECUNDS) r 1 LEV iL(INCH-REF.SEP-$MRT) ! VOI ) REACTIVITY 2 VESSEL STEAMFLOW 2 DOP)LER REACT :TY 3 TUR3fNE STEAMFLOV r 3.0 3, SCR, 19 ,L LMegactru REACTIV L
' rEE JunTEn rLCu 150.0
- . 0 : ; .
\ O
\
a
--- {
- e. :
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i,
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h 8
N P O
=
50.0 - .l.0 f I 1
m
{I l
- 0. 0 2.O
- 0. 3 10.0 20.0 0. 0 10.0 28.0 TIME (SECONOS) TIME (SECONDS)
E E
Figure 4a. Plant Response to Feedwater Controller Failure (EOC) 7 16
.h --
23A1806 R:v. 0 150.0
- NEUTRON FLUX 1 VESSEL PRESS RISE S3 2 SAF ETY VALVE FLOW 2 AVE SURFACE HEAR FLUX 3 COR E INLET FLOW i 3 REL [EF VALVE FLOW
' Ean " E' E'? 4 BYPNSS VALVE FLOW
'150.0 100.0
\
P i g
v v.- -
y 100.0 ; ; ;
a b 30.0
+
.5 .
.W t' a s ^ '
y -
30.0 M
0.0 , M,. . , . . , _ .
,. . L. . .-
0.0 20.0 0. 0 10.0 20.0
- 0. 0 10.0 TIME (SECONOS1 TIME (SECONOS) i 1 LEV EL(INCH-REF-SEP-SKRT) i VOI ) REACTIVITY 2 VES iEL STEAMFLOW 2 00P'LER REACTiv ; f r
3 TURllNE STEAMFLOW 3 SCRkPt erarv,ure REACTIVIT 1.8
- int u_ r 150.0 ' rEE ?witEn etnu 0 . ?
)2 ) " il 0.0 m m_
_t i, 100.0 w . . .
(- '
B \y
) :
50.0 _[ h-1.0
/
O. 0 _ .. 0 o 20.0 0. 0 10.0 20.0
- 0. 0 10.0 TIME (SECON05)
TIME (SECON05)
Figure 4b. Plant Response to Feedwater Controller Failure (EOC-1000 mwd /ST) 17
m'
- 23A1806 Rev. 0 I
150.0 1NEUIRON FLUX 1 VES SEL PRESS RISE P 13
- 2 AVE SURFACE HE LT FLUX 2 SAF ETY VALVE FLO' 3 COR E INLET FLCi 3 REL lEF VALVE FLO 150.0 ' cae r !utET Ego ,
4 BYP\SS VALVE FLC 300.0 l
'E # : 2 -
-}' l 0 100.0 ;- ; ; ; *
= a B d
_ s0. 0 l
l M
- e e . .
" 50.0 I C '
]
- 0. 0
,, , , . _ , _ , , , .4 . ,
e E
10.0 20.0 0. 0 10.0 20.0
- 0. 0 TIME (SECONDS) TIME (SECONOSI
. i VOI ) REACTIVIT'r I 1 LEV EL(INCH REF-SEP-SKRT) 2 VES3EL STEAMFLOW 2 DOP)LER REACTI / TV a rY E
3 TUR 31NE STEAMFLOW 1.0 3,SCR 193 LM itRE ecA.CT,I,VI 7 y3 rv
- 150.0 e rEE JuaTEn rLey
, 4
= :
\
~
) C I :: 0 : O C U Ch % h 0. 0 . , _ + _ _, m . -
=
w
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-s 8 ,
.a h
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- j i
s 0. 0 -# t -
t 9
w
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u "
f 0. 0 , - -2.0 20.0 0. 0 10.0 20.C
- c. '1 10.0
. TIME (SECONOS) !!ME (SECONOS)
L i
e B
I Figure 4c. Plant Response to Feedwater Controller Failure (EOC-2000 mwd /ST) 18
23A1806 Rev. 0 02 06 10 14 18 22 26 30 34 38 42 46 50 51 14 14 14 47 14 12 12 14 43 '.4 36 36 36 14 39 8 4 0 0 4 8 35 8 14 36 36 36 14 8 31 26 6 4 4 6 26 27 10 8 14 14 14 8 10 23 26 6 4 4 6 26 19 8 14 36 36 36 14 8 15 8 4 0 0 4 8 11 14 36 36 36 14 07 14 12 12 14 03 14 14 14 i
NOTES: 1. No. indicates number of notches withdrawn out of 48. Blank is a withdrawn rod.
- 2. Error rod is (22,39).
Figure 5. Limiting RWE Rod Pattern i
19
_. _ _ _ _ . _ . ' ~ - " "
E 2 l 23A1806 Rev. O L
L.
' 1 NEUTRON F _UX 1 VL55EL PR[ SS R]SE LPSI) 2 AVE SLRF ATE HE A T FLUX 2 SAFE TY WAL VE FLOW l
b i E
I S ,
W 100.0 ,' 200.0 ,
= i h N # N L b F
s- ,
5 ti r
~ 5 0. 0 100.0 p _ _ _ _ _ .
. _ _ . _?-
_ _ _ _3 3 3 3 4 KI F
=
f--
'r . , ' . . : : : ;
ar 0. 0 5.0 0.0 5.0 -
7 TIME (SECONOS) Tlei (SE CONOS) i W
h ,
v 5 1 LE VEL (INC4 REF -SEP-SKRT) l VOID REACT!vlTV j 2 VESSEL STEAMFLOW DOPPLER REAlily!TV r 3 TURBINE STEAMFLOW 3s AM REAETIVITY 200.0 gerrnwavro eggy 1.0 4g .ggggy i
[\/ -
H W h, 100 0 3 0. 0 .. ,, ,
- ym ;__ _ __ g m % 8 L
r s-
- 8. 0 m , , , -1.0
- y a a a a y
=-
-100.0 -2.0 L
- 0. 0 5.0 0. 0 5. 0 b T IME (SECONDS) TIME (SECONDS) g ..
W
[
a
. Figure 6. Plant Response to MSIV Closure (Flux Scram) r-20
+
23A1806 Rev. O A NATURAL CIRCULATI ON 9105 PER':ENT ROD L INE C ULT. PERFORMANCE LIMIT 1.00 A
G .75 N
N X
C$
" .50
/
tt a
.25
\ \s n 0.00 *"
- 0. 0 20.0 40.0 60.0 80.0 100.0 120.&
PERCENT POWER Figure 7. Reactor Core Decay Ratio versus Power 21
23A1806 Rev. 0
- 0. 0
-5.0
/
T -10.O jd
,.n
- uggo,. . n s ,
aq. e. ,
z N w -l1'Q"Q '
~
o -15.0 u
/ ;'n' '*.
p .
e
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w w .e.: -
/;W*
o u ;;3c -9.
c: -20.0 'i,. ',. . T,t w _, , r. . .
a
- a. n..
O. ;. 'O: i y o ? "g~~
o '.,e...;
.c.
-25.0 V ,.5_.
' :Sa-'4 g
"'e -
A CALCUL STED VALUE- COLD -
8CALCUL NTED VALUE- HSB '$ 9 :.1.4 C BOUND /AL 280 cal /G COLD Kfi .,
D BOUND /AL 280 cal /G HSB 4 ' . .- -
-30,0 l:Q .?
4
. p t
- 0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.O- ) 4P.r.(l .h FUEL TEMPERATURE DEG C. i..:
a.
Figure 8. Doppler Reactivity Coefficient in 1/4 *C 22
I 23A1806 Rev. O E
i u
20.0
- 17.5
! 15.0 E
m E ca 12.5 #U U x x gj 10.0 s
e r >-
/
" C 7.5
+
h o
< a h & 5.0 i w
r a
E 2.5 E
T A ACCIDENT FilNCTION k B BOUNDING VALUE 280 CAL /G
- 0. 0
_ 0. 0 5.0 10.0 15.0 20.0 5 ROD POSITION, FEET OUT g
Figure 9. Accident Reactivity Shape Function, Cold Startup
[
6 23
F 23A1806 Rev. O i
g 20.0 I
L
! 17.5 E
F
$ 15.0 a
! I w
6 + 12.5 .
g-e-e-- ,.
a <
P H b gj 10.0 ,
E o e
& C 7.5 E 3 h 0 e
@ 5.0
=
a
=
r 2.5 E
A ACCIDENT FilNCTION p B BOUNDING VaLUE 280 CAL /G
_ 0. 0 7
k 0. 0 5.0 10.0 15.0 20.O ROD POSITION, FEET OUT f
m
[
R h
$ Figure 10. Accident Reactivity Shape Function, Hot Standby
?
k 24 L
E-
23A1806 Rev. 0 1
4 4
40.O A SCRAM F JNCTION
- B BOUNDIN 3 VALUE 28 0 CAL /G 35.0 ,
i m 30.0 i / 9 w
- p 25.0 .
N ci a
,., 20.0 -
e w
5 g 15.0 s
G C
10.0 u
5.0
- 0. 0 ,., ,., / , __n___.-t m -
- 0. 0 1.0 2. 0 3.0 4.0 5.0 6. 0 ELAPSED TIME, SECONDS l
1~~.
Figure 11. Scram Reactivity Function, Cold Startup ',
25
23A1806 Rev. 0 60.O __
A SCRAM F JNCTION '
8 BOUNDIN 3 VALUE 281) CAL /G 50.0 A . .
a 40.0 x
5 Ei a
m 30.0 0
5 b 20.0 a E
L; 5
x 10.0
- 0. 0 b\
/
&b d
/
, , _ - L
- 0. 0 1.0 2. 0 3.0 4.0 5.0 6. 0 ELAPSED TIME, SECONDS Figure 12. Scram Reactivity Function, Hot Standby 26
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23A1806 Rev. 0 .; , - - '
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f APPENDIX A t /. ^ -
GETAB ANALYSIS INITIAL CONDITIONS . /I. . .
<L,
- . .~y .,
t'-,,
..u, % ,:-
j The values listed below were used in the GETAB analysis for this reload ,...'..,
- g rather than the values given in Reference A-1, to more nearly reflect actual . . . ,i}.[ #
's .
'.1-' plant data. . . . . .
.7 ,
g +
s.
- v. :
1 5
0 . .
's-2- Reactor Pressure (psia) 1035 y .....,--
x9-,-
Inlet Enthalpy (Btu /lb) 527.0 . J.< L .-
~
- .=,,.,._
ff .
3 O REFERENCE ~l " m- 5 -
.-s.-
...g. , ,e ,. eg
- w-
- I -.
.ku . .
A-1. " General Electric Standard Application for Reactor Fuel,"
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. l ; q.
' gj_.,
NEDE-24011-P-A-6, April 1983. .-
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I GEN ER AL $ ELECTRIC m,
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