ML19323G365
| ML19323G365 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/31/1980 |
| From: | Engel R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19323G359 | List: |
| References | |
| 80NED258, NEDO-24239, NUDOCS 8006020243 | |
| Download: ML19323G365 (32) | |
Text
NEDO-24239 80NED258 Class I f) b January 1980 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR i
BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 RELOAD 2 l
l r
Approved: M' ' -
R. E. Enges Manager Reload Fuel Licensing l
I i
i NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC 9 0 0 (p 0 M 0 4 0 l
r NEDO-24239 INPORTANT 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 Commission (USNRC) for amending CP&L's operating license of the Brunswick Steam Electric Plant Unit 1. The information contained in this report is believed by General Electric to be an accurate and true representation of l 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 Fuel Contract Supplemental Agreement between Carolina Power and Light and General Electric Company for Brunswick 1 and 2, dated January 28, 1974, 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 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. PLANT-UNIQUE ITEMS (1.0)*
New Bundle Loading Error Analyses Procedures: Appendix A i Densification Power Spiking: Appendix B Transient Operating Parameters: Appendix C i 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Type Number Number Drilled Irradiated Initial Core 228 228 Reload 1 8DRB265L 52 52 Reload 1 8DRB283 124 124 New Reload 2 P8DRB265H 16 16 i
Reload 2 P8DBR285** 140 140 Total 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure: 12,060 mwd /t Assumed reload cycle exposure: 14,260 mwd /t Core loading pattern: Figure 1
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH -
NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)
BOC k gg Uncontrolled 1.094 Fully Controlled 0.938 Strongest Control Rod Out 0.972 r
R, Maximum Increase in Cold Core Reactivity-with Exposure Into Cycle, Ak 0.008
- ( ) refers to areas of discussion in Generic Reload Fuel Application, NEDE-240ll-P-A-1, August 1979.
- Letter, R.E..Engel to T.A. Ippolito, " General Electric Licensing Topical Report NEDE-240ll-P-A, General Electric Application, Appendix D Submittal,"
December 14, 1979.
1
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) ppm (20*C, Xenon Free) 600 0.045
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)
E0C3-2000 BOC3 to mwd /t to EOC3-2000
' EOC3 mwd /t Void Coefficient -8,50/-10.62 -8.70/-10.87 N/A* (c/% Rg) 41.6 41.6 Void Fraction (%)
-0.225/-0.214 -0.220/-0.209 Doppler Coefficient N/A (c/*F) 1374 1374 Average Fuel Temperature (*F)
-37.93/-30.34 -36.72/-29.38 ,
Scram Worth N/A (S)
Figure 2a Figure 2b Scram Reactivity
- 7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAME 8x8R P8x8R 8x8 BOC3 to BOC3 to .
BCC3.to E0C3- E0C3-2000 EOC3- EOC3-2000 E0C3-E0C3-2000 2000 ' mwd /t to 2000 mwd /t to 2000 mwd /t to mwd /t EOC3 mwd /t EOC3 mwd /t E0C3 1.20, 1.20, 1.20, 1.20, Peaking factor 1.22, 1.22, 1.47, 1.54, 1.45, 1.52, (local, radial, 1.34, 1.40, 1.40 1.40 1.40 1.40 1.40 axial) 1.40 1.051 1.051 1.051 1.098 1.098 R factor 1.051 5.960 6.26 6.558 6.178 6.487 Bundle Power (MWt) 5.703 112.1 114.3 112.1 115.4 113.0 Bundle Flow 114.2 (103 lb/hr) 1.24 1.30 1.24 1.32 1.25 Initial MCPR 1.30 ,
~N = Nuclear input data
~A = Used in transient analysis 2-
NED0-24239
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2.2)
Thermal Power Monitor
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Core I
Power Flow $ Q/A SL V ACPR Flant Transient Exposure (%) ] (Z NBR) (% NBR) psig psig 8x8/8x8R/P8x8R Response Load Rejection EOC3-2000 104 100 307 117 1174 1213 0.23/0.23/0.25 Figure 3t/a No Bypass Wd/t to EOC3 Load Rejection BOC3 to 104 100 275 114 1169 1210 0.17/0.17/0.18 Figure 3b No Bypass EOC3-2000 WJ/t u se of 100*F BOC3 to 104 100 125 121 1023 1068 0.14/0.15/0.15 Figure 4 Feedwater Heater EOC3 Feedwater EOC3-2000 104 100 187 113 1146 1187 0.14/0.14/0.15 Figure 5a Controller Failure Wd/ t to EOC3 Feedwater BOC3 to 104 100 163 111 1143 1184 0.10/0.09/0.10 Fir se 5b Controller Failure EOC3-2000 Wd/t
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(5.2.1)
ACPR LHGR Rod Block Rod Position 8x8/P8x8R 8x8/P8x8R Limiting Reading (Feet Withdrawn) and 8x8R and 8x8R Rod Pattern 104 4.0 0.11/0.10 11.3/13.7 Figure 6 105 4.5 0.12/0.10 11.3/13.7 Figure 6 106 5.5 0.14/0.12 11.3/13.7 Figure 6 107* 6.0 0.15/0.13 11.4/13.6 Figure 6 108 9.5 0.18/0.15 -
11.0/15.0 Figure 6 109 10.5 0.18/0.15 11.1/15.4 Figure 6
- Indicates setpoint selected 3
L _ ..
.y ._ ,
^
- 11. OPERATING MCPR' LIMIT (5.2, Appendix C) f BOC3 to EOC3- E0C3-2000 mwd /t 2000' mwd /t. to EOC3 mwd /t 1.24 1.30 (8x8 fuel) 1.24 1.30 .(8x8R fuel) 1.30 1.32 (P8x8R fuel) i 12 OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Flow s1 y Plant Transient (%) (%) (psig)_ (psig) Response 4 MSIV Closure 104 100 1222 1252 Figure 7 (Flux Scram)
- 13. STABILITY ANALYSIS RESULTS (5.4)
! Decay Ratio: Figure 8
- Reactor Core Stability:
Decay Ratio, x2/*o 0.74 (Natural Circulation-105% Rod Line) ,
i Channel Hydrodynamic Performance Decay Ratio
'(Natural: Circulation-105% Rod Line)
- 8x8 Channel 0.57 8x8R Channel 0.47
. P8x8R Channel- 0.47, i
4.
7s.
U" s t {.. .
i . ,. L ,. .-
- 14. ' LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
( INIT. CORE, TYPE 1 Local Exposure MAPLHCR P.C.T. Oxidation (mwd /t) (kW/ft) (*F) Fraction 200 11.7 2113 0.022 1,000 11.8 2113 0.022 5,000 12.0 2128 0.023 10,000 12.1 2126 0.023 15,000 12.3 2166 0.026 20,000 12.0 2141 0.024
- 25,000 11.1 2014 0.016.
30,000 10.1 1376 0.009 I
INIT. CORE, TYPE 2 4
200 11.0 2017 0.018 1,000 11.1 2016 0.018-5,000 11.7 '2078 0.020 10,000 12.2 2146 0.024 15,000 12.2 2163 0.026 20,000 12.0 2149 0.025 3
25,000 11.1 2026 .0.017 30,000 '10.1 1886' O.010
.I
. S o 1
NED0-24239
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2) (Continued)
Fl?EL 8DRB265L Local Exposure MAPLHGR P.C.T. Oxidation (mwd /t) (kW/ft) (*F) Fraction 200 11.6 2128 0.026 1,000 11.6 2129 0.025 ,
5,000 12.1 2178 0.029
! 10,000 12.1 2169 'O.028 15,000 12.1 2183 0.029 20,000 11.9 2170 0.029 25,000 11.3 2101 0.023 30,000 10.7 2020 0.017 FUEL 8DRB283 200 11.2 2090 0.023 1,000 11.2 2083 0.022 5,000 11.8 2149 0.027 10,000 12.0 2161 0.028 15,000 12.1 2180 0.029 20,'000 11.8 2164 0.028 25,000 11.3 2096 0.023 30,000 11.1 2072 0.021 .
e 6
- r'
. 14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2) (Continued)
P8DRB235 Local E:aposure MAPLHGR P.C.T. Oxidation (mwd /t) (kW/ft) (*F) Fraction 200 10.9 2038 0.019 1,000 11.0 2048 0.020 5,000 11.8 2141 0.026 10,000 12.3~ 2177 0.029 15,000 12.2 2174 0.028 20,000 11.8 2131 0.025 25,000 11.0 2031 0.018 30,000 10.4 1928 0.012 i
i
! P8DRB265H 200 11.5 2103 0.024 I 1,000 11.6 2111 0.024 ,
5,000 11.9 2135 0.025 10,000 12.1 2147 0.026 15,000 12.1 2157 0.027 20,000 11.9 2138 0.025 25,000 11.3 ,2063 0.020
'30,000 10.7 1977 0.015 L
9
~
1 t
- 7
- i .L
- 15. LOADING ERROR RESULTS (5.5.4)
Limiting Event: Rotated P8DRB285 MCPR: 1.07 16 CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficients: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 8
WEDO-24239 -
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= 88888888888i8888888 m 8 8 _8 8i8 8 8 8 8 8 8 8 o2 8 8 8'8 8 8 8 8 8 8 i i ll IIIIIII 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 FUEL TYPE A = INITIAL CORE TYPE 1 E = RELOAD 2 PBDRB265 8
- INITIAL CORE TYPE 2 F = RELOAD 2 PBDRS285 C
- RELOAD 180R8266 G=
D = RELOAD 180RC283 H=
Figure 1. Reference Core Loading Pattern 9
m.
NEDO-24239 100 45 90 -
1 80 -
35 678 CRD IN PERCENT 7t g -
2 NOMIN AL SCR AM CURVE IN (-8) 30 60 -
SCRAM CURVE USED IN ANALYSIS b - 3 GJ.
z 9 >
2 a b se -
2 o
5
- 20 40 -
15 30 -
10 20 -
l l
1 5
10 1
2 0= 0 0 1 2 3 4 TIME (sec)
' Figure 2a. Scram Reactivity and Control Rod Drive Specification l EOC3-2000 mwd /t to E0C3 10 l
NEDO-24239 100 46
- - _ e 1
80 -
878 CRD IN PERCENT _ g 70 - NOMIN AL SCR AM CURVE IN (-$)
- so 2
SCR AM CURVE USED IN ANALYSIS eo -
a
_I S so _ $-
5 2:
E b E
4o -
- ts 30 -
- 10 20 -
1 - 6 10 -
2 I
0 O O 1 2 3 4 TIME (ese)
Figure 2b. Scram Reactivity and Control Rod Drive Specification BOC3 to EOC3-2000 l
11 l
1' ! NEUTRCN FLUX X VESSEL FfkS RISE (PSI) 2 AVE SURFFCE HEAT FLUX 2 SAF ETY V9t.VE FLOW 150. 3 CORE INLE T FLCW 300* 3 PELIFF V R.VE F (H_
4 4 BIPHSS VN_'VIT(LLN i S S 6
gi. s m.
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i 50. MA f '
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- 0. 4. 8. 12. 16. O. 14 . 8. 12. 16.
2 B
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u H
N I VOID KFETIVITY 1 LEVEL (INCH-E F-SEP-SKIRT 2 VESSEL SlEAMFLON 2 DOPPLER l-EACTIVITY C
w 3 TUR8INE S TEAMFLOW g* 3 SCRAM REECTIV!TT e 200*
4 FEEDWHTEF FLOW 4 TOTAL REfCTIVII)
S a
100' '" - * $ 0* 3 '
N l2 N \
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- 0. i 3 3 u ,_ .g, 3 i
- g W
N 100. - ' - * -2. ' -- -
O. 4 8. 12. 16. O. 0.4 0.8 1.2 1.6 TIE (SEC) TIME ISEC)
Figure 3a. Plant Response to Generator Load Rejection, Without Bypass, End of Cycle 3 - 2000 mwd /t to End of Cycle 3
1 ' 11 VEcc,EL FsES RISE !P31)
I 1 NEUTRON FLUX 2 AVE SURFFCE HERT FLUX 2 x tri vaE rlow 3 M IUT FLW 31._Lj[r vi t vf_F(pW_
150, 300* 4 b 4A55 4 G E fLC*
4
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- 0. 16. O. 4. 8. 12. 15.
.O. 4. B. 12. TIME ISEcl TIME (SEC1 h
8 a
1 VOID KXTIVITY $
w 1 LEVEL (INCH-REF-SEP-SKIRT 2 DOFTLEH FfRCTIVITT 2 VESSEL S1EAMFLOM 'O 3 SCRAM RErCTIyJTT 3 TURBINE 5 TEAMFLOW 3* / lit E TUCTTiiT Y '"-
200* 4 FLEDWATEF FLOW S
E C 0* 3 " /2 \
100* -
- 2. < 5 M( %Q .,
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- 0. 4. a. 12. 16.
TIME ISEC1 TIME (TCl
- Figure 3b. Plant Response to Generator Load Rejection, Without Bypass, Beginning of Cycle 3 to End of Cycle 3 - 2000 mwd /t
2M8Ls~ $)
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a -
3 CORE INLEi FLOW 125' y BiPh5S VFLVE FLdd 150. 9 CORE int.t T SUB S
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- 40. O. 10. 20
- 00. 10. 20 30.
=
8 JO I VOID WT!VITT u l 1 LEVEL (IMH-REF-SEP-SKIRT C' 2 00PP], R FEACTIV 2 VESSEL SlEAMFLOW 3 Staph RTECT IVyff 3 TURBINE 5 TEAMFLOW g* DL REfCT WiTY 150* . 4 q_
4 FEEDWRTEF FLOW t 5
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- 30. 14 0 .
4 O. 6. 12.
TIME (SECl
- 18. 24.
TIME (SEC1 Figure Sa. Plant Response to Feedwater Controller Failure, End of Cycle 3 -
2000 mwd /t to End of Cycle 3
1 EUTRON FLUX 1VESSELPdESRISE(PSil
( 2 AVE SURFFCE HERT FLUX 2 SAF E T Y VitVE FLt%4 3 CORE INLE T FLOW 3.RELI Q,,)PJE Ftpl_
150, 125* f 4 BrfHn vr_vE"TLOW 4 CORT !NLt T SUS 5 5 1
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- 30. 40. O. 10. 20. 30.
- 0. 10. 20. TIME [SEC)
TIME ISEC) h 8
5 k I LEVEL (INDf-REF-SEP-SK1 FIT / 1 VOID RE#TIVITy'TY $
I 2 /EACT4VI 2 VESSEL S1EAMFLOW 3 TURBINE ETEAMFLOW 3 SC REfCil1/ITY 3*
150*
q_
FEEDWRTEF FLOW \ 4 (CTiET DIITY O.
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w kr
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- 0. ba.ud.uu 3 3u 3 a 2. '-
24.
- 40. D. 6. 12. 16.
O. 10. 20. 30. TIME (SEC1 TIME ISECl
~
Figure 3b. Plant Response to Feedwater Controller Failure, Beginning of Cycle 3 to End of Cycle 3 - 2000 mwd /t
t
- 02 06 10 14 18 22 26 00 51 38 47 43 36 00 30 39 14 10 35 38 18 00
=
31 00 27 00 10 00 i
E Notes: 1. Rod Pattern Is 1/4 Core Mirror Symmetric, j- -Upper Left Quadrant Shown on Nbp.
- 2. Numbers Indicate Number of Notches Withdrawn out of 48. -Blank Is a Withdrawn Rod.
- 3. Error Rod Is (18-27).
4 1
Figure 6. . Limiting RWE Rod Pattern e
17
1 '
1 NEUTRON FLUX 'l VESSEL FFES RISE (PSI) 2 AVE SURFFCE HEAT FLUX 2 SAFE T T VfLVE FLOW 150. LCORE INLET FLOW 300. 3 8fLI E VfO LFLUN-4 9 bifMS 'QL FLO.i S 4 5 Ny' 100. 200. A i N b
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TIME ISEC) TIME (SEC)
~
2:
9 ~
~
?
N 1 LEVEL (IPKH-REF-SEP-SKIRT 1 VOID REfEJD/rfi U 2 VESSEL SlEAMFLOW 2 00PFLFWTEACT1VITY 200* 3 TURBINE E TEAMFLOW 3* 3 SCRnN REfCTivirr 4 FEEDWRTEF FLOW 4 IAL REffIlVITY S
100. J ? , , - - D. B --
k N s
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D. _
2
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-100. ' - * -2. - t- 4 D. 4. 8. 12. 16. O. 0.6 1.2 1.8 2.4 TIME (SEC) TIME (SEC) u Figure 7. Plant Response to MSIV Closure
NEDO-24239 1.4 1.2l-ULTIMATE PERFOttMANCE LIMIT 1.0 - - - - - - - = = - - - - - - - = = = = = - - - - = = = = - = = =
, 02 -
X
'n h NATURAL p CIRCULATION C
S 0.6 -
0 106% ROD UNE 0.4 -
0.2 -
0.0 0 20 40 00 80 100 120 PERCENT POWER Figure 8. Decay Ratio 19 l
1
-10 -
O
-15 -
z 9
i I '
E o
g -20 -
B O
-25 -
l
[.3 BOUNDING VALUE FOR 280 cal /g COLD
~
8 BOUNDING VALUE FOR 280 cal /g HSB O CALCULATED VALUE - COLD 9 CALCULATED VALUE-HSB
-36 I I I I I 4
O 400 800 1200 1600 2000 2400 FUEL TEMPERATURE ('C) l Figure 9. Doppler Reactivity Coefficient Comparison for RDA l 20 l
NEDO-24239 24 0 BOUNDING VALUE FOR 280 cal /g 8 CALCULATED VALUE 20 -
16 -
E E
5 a
D k
w > -d "1, 12 -
C 5
P E
E 8 -
4 -
I I I I og 16 20 O 4 8 12 i
ROD POSITION (ft OUT) ,
i Figure 10. RDA Reactivity Shape Function at 20*C 21 l
NEDO-24239 24 O SOUNDING VALUE FOR 280 cal /g 8 CALCULATED VALUE ,
20 -
16 -
a i
5 0
x C g t- 12 -
l 2
D 5
G N
E 8 -
_^ r--a 4 - 2) l l g I I OL 4 8 12 16 20 0
ROD POSITION (ft OUT) l l
l Figure 11. RDA Reactivity Shape Function at 286*C 22
NED0-24239 so l
O BOUNDING VALUE FOR 280 cal /g 70 - O CALCULATED VAlt;E 00 -
l l
I E" -
a z
$3 O
h
[4 40 -
I 5:
\
l 4 .
1 Ex -
20 -
7
- 3---
1 i
i 10 l
1 I
Of O 2 4 6 8 to ELAPSED TIME (sec) i l
Figure 12 RDA Scram Reactivity Function at 20*C 23
NEDO-24239 120 -
l 1
i O souen NG VALUE FOR 280 cal /g
,g _ d CALCULATED V ALUE I
l 1
l l
i l
Bo -
6 O
k 5
o l
l .
w -
- 7. eo D
5 P
N e
4o -
20 -
O D OL 0 2 4 6 9 to ELAPSED TIME (sec)
Figure 13. RDA Reactivity Shape Function at 286*C 24
NEDO-24239 APPENDIX A NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES The bundle loading error analyses results are based on new analyses procedures for both the rotated bundle and the miolocated bundle loading error events.
The use of these new analyses procedures is discussed below.
A.1 NEW ANALYSIS PROCEDURE FOR THE ROTATED BI!NDLE LOADING ERROR EVENT The rotated bundle loading error event analysis results presented in this supplement are based on the new analysis procedure described and approved in Reference A-l. This new m,ethod of performing the analysiu is based on a more accurate detailed analytical model.
The principle difference between the previous analysis procedure and the new analysis procedure is the modeling of the water gap along the axial length of the bundle. The previous analysis used a uniform water gap, wheteas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interfacing between the top guide and the fuel spacer buttons, caused by misorientation, causes the bundle to lean.
The effect of the variable water gap is to reduce the power peaking and the s
R-factor in the upper regions of the limiting fuel rod. This results in the calcu'.ation of a reduced CPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry.
e The results of the analysis indicate for the P8DRB285 bundle a 17.7 kW/ft LHGR (includes densification spiking penalty of 2.2%) and 0.23 ACPR (includes
' a 0.02 penalty due to variable gap R-factor uncertainty) with a CPR of 1.07.
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.i LA. 2 NEW ANALYSIS PROCEDURE FOR Tile 'MISLOCATED BUNDLE LOADING ERROR EVENT
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h The .mislocate'd bundle loading error event analyses results presented in this supplement are based on the new analysis procedure described in Reference A-1.
This new method of performing the analysis employs a statistically corrected llaling procedure and analyzes every bundle in the core.
The use of the statistically corrected Ilaling analyses' procedure indicates
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that the minimum CPR for mislocated bundles (e.g., P8DRB26511 into Initial / Core) is greater than the safety limit (1.07) for all exposures throughout the cycle.
i The linear heat generation rate is less than that of the rotated bundle.
i-REFERENCES A-1 Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engle (GE),
MFN-200-78, dated May 8, 1978.
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NEDO-24239 APPENDIX B DENSIFICATION POWER SPIKING Reference B-1 documents the NRC staff position that ". . . it (is) acceptable to remcVe the 8x8 and 8x8R spiking penalty factor from the plant Technical Specification for those operating BWR's for which it can be shown that the predicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGR's."
The BSEP-1 Reload-2 submittal contains the required information to demon-Section 10, strate that the stated criterion is met for BSEP-1, Reload 2.
Rod Withdrawal Error, and Appendix A (New Bundle Loading Error Event Analysis Procedures) include the densification effect in the calculated LHGR of the 8x8 fuels.
REFERENCE B-1 " Safety Evaluation of the General Electric Methods for the Consideration ,
of Power Spiking Due to Densification Effects in BWR 8x8 Fuel Design and Performance," Reactor Safety Branch, DOR, May 1978.
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NEDO-24239 APPENDIX C TRANSIENT ANALYSIS INITIAL CONDITIONS S/RV Lowest Setpoint (psig) 1105 + 1%
S/RV Capacity (%) 86.9 l
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