ML19242B441
| ML19242B441 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/30/1979 |
| From: | Brugge R, Ervin A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19242B438 | List: |
| References | |
| NEDO-24199, NUDOCS 7908080524 | |
| Download: ML19242B441 (20) | |
Text
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aun^fo*72 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY NUCLEAR POWER STATION UNIT 3 RELOAD NO. 2 509 184 GENER AL h ELECTR 7 90808 0 l:l ';
/'~
NEDO-24199 79NED281 Class I June 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY NUCLEAR POWER STATION UNIT 3 RELOAD NO. 2 Prepared:
M N
A. M.
Ervin N. k,h' Approved:
R.
O.
rugge, Manager Operating Licenses II NUCLE AR ENERGY P;iOJECTS DIVISION
NEDO-24199 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for The Tennessee Valley Authority (TVA) for TVA's use with the U.
S. Nuclear Regulatory Commission (USNRC) for amending TVA's operating license of the Browns Ferry Nuclear Unit 3.
The information contained in ti.is 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 t.ndertakings of the General Electric Company respecting information in this documei.t are contained in the contract between The Tennessee Valley Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Browns Ferry Nuclear Plant Unit 3, dated.Iune 17, 1966, and nothing contained in this document shall be construed as chant;ing 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 useful-ness of the information contained in this document or that such use of such information may not infringe privately owned rights; ar do they assume any responsibility f or lio:lity or damage of any kind which may result from such use of such infornation.
509 W6
PLANT-UNIQUE ITEMS (1.0)*
Items different from or not included in Reference 1:
Fuel Loading Error LHGR: Appendix A Safety / Relief Valve Capacity: Appendix A Spring Safety Valve Capacity:
Appendix A new Bundle Loading Error Eu nt Analysis Procedures: Re ference 3 LHGR includes 0.02 penalty fc _ R-factor uncertainty 2.
RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0)
Fuel Type Number Number Drilled Irradiated 8DB219-Initial Core Central 320 320 Irradiated 8DB219-Initial Core Peripheral 92 92 Irradiated 8DRB265L 208 208 New P6DRB263L 144 144 Total 764 764 3.
REFERENCE CORf' LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycle:
12,686 W d/t Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:
12,486 Wd/t Assumed reload cycle core average exposure at end of cycle:
14,040 5%'d / t Core leading pattern:
Figure 1 4
CALCULATED CORE EFFECTIVE MI'LTIPLICATION AND CORE SYSTEM WORTH - NO VOIDS, 20"C (3.3.2.1.1 AND 3.3.2.1.2)
BOC keft.
Uncontrolled 1.108 Fully Controlled 0.949 Strongest Control Rod Out 0.989 R, >!a x imum Increase in Cold Core Reactivity with Exposure Into Cycle, ak 0.000
- ( ) Refers to areas of discussion in " Generic Reload Ftv l Application,"
NEDE-240ll-P-A, Revision 0, August 1978.
509 187 1
NEDO-24199 5.
STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) 600 ppm Shutdown Margin (t.k) 0.040 (20 C, Xenon Free) 6.
RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)
Void Coefficient N/A* (c/% Rg)
-6. 44/-8.43 Void Fraction ( 7. )
40.29 Doppler Coefficient N/A (c/ F)
-0.228/-0.217 Average Fuel Temperature ( F) 1337 Scram Worth N/A ($)
-37.713/-30.17 Scram Reactivity vs Time Figure 2 7.
RELOAD-UNIOUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) g osure 8x8 8x8R P8x8R_
Peaking factors 1.22 1.20 1.20 (local, radial 1.4472 1.5858 1.5309 and axial) 1.40 1.40 1.40 P-Factor 1.098 1.051 1.051 Bundle Power 6.104 6.681 6. 6 6')
(MWL)
Bundle Flow 108.2 108.6 109.0 (103 !b/hr)
Initial MCPR 1.22 1.22 1.23
- N = Nuclear Input Data A = Used in Transient Analysis 509 1E8 2
NEDO-24199 d.
SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
Recirculation Pump Trip 9.
CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.2)
Core p
p Power Flow Q/A s1 v
ACPR Plant Transient Fxposure
(%;
g (I NBR)
(I NBR)
(psig)
(psig) 8xR 3x8R P8x8R
Response
Load Rejection BOC-EOC 104.5 100 229.5 109.3 1206 1232 0.15 0.15 0.16 Figure 3 without Bypass Loss of 100*F 104.5 100 123 122.06 1012 1068 0.14 0.14 0.14 Figure 4 Feedwater Heating Feedwater BOC-EOC 104.5 100 159.8 110.7 1154 1188 0.10 0.10 0.10 Figure 5 Controller Failure 10.
LOCAL ROD WITilDRAWAL ERROR (WITH LIMITING INSTRUMEFT FAILURE)
SUMMARY
(5.2.1)
Rod Position Rod Illock (Feet ACPR MLHGR Limiting Re.iding Withdrawn) 8x8 8xfd P8x8R 8x8 8x8R P8x8R Rod Pattern 104 3.5 0.16 0.11 0.09 14.6 16.4 14.1 Figure 6 105 4.0 0.19 0.13 0.11 15.4 16.8 14.5 Figure 6 106*
4.5 0.21 0.14 0.13 15.5 17.0 14.6 Figure 6 107 5.0 0.23 0.16 0.14 15.3 17.0 14.7 Figure 6 108 5.5 0.25 0.17 0.15 15.1 16.8 14.7 Figure 6 109 6.0 0.26 0.18 0.17 14.9 16.6 14.7 Figure 6
- Indicates setpoint selected 509 3
NEDO-24199 11.
OPERATING MCPR LIMIT (5.2)
F20C3 to EOC3 1.28 8x8 Fuel 1.22 8x8R Fuel 1.23 P8X8R Fuel 12.
OVERPRESSURIZATION ANALYSIS SUSD1ARY (5.3)
Pover Core Flow sl v
Plant T_ransient
(%)
(%)
(psig)
(psig)
Response
MSIV Closure 104.5 100 1246 1280 Figure 7 (Flux Scram) 13.
STABILITY ANALYS15 RESULTS (5.4)
Decay Ratio:
Figure 8 Reactor Core Stability:
Decay Ratio, x,/x 0
(105% Rod Line - Natural Circulation Pover)
Channel Hydrodynamic Performance Decay Ratio, x /x0 (105% Rod Iine - Natural Circulation Power) 8x8R/P8x8R 0.273 8x8 0.383 14.
LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
Reference 2.
509 i90 4
NEDO-24199 15.
I.0ADING ERROR RESUI.TS (5.5.4)
Limiting Event:
Rotated Bundle 8DPR265L MCPR: 21.07 16.
CONTROL R0D DROP ANi LYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient:
Figure 9 Accident Renetivity Shape Functions:
Figures 10 and 11 Scram Reactivity Functions:
Figures 12 and 13 Plant specific analysis results Parameter not bounded: Accident Reactivity Shape Function at 20 C Resultant peak enthalpy:
278 cal /gm 509 191 5
NEDO-24199 REFERETCES l.
General Electric lloiling Water Generic Reload Fuel Application, NEDE-240ll-P-A, August 1978.
2.
Loss-Of-Coolant Accident Analysis Report for llrowns Ferry Nuclear Plant Unit 3, NEDO-24194, June 1979.
3.
Supplemental Reload I.icensing Submit tal f or 15rowns Ferry Nuclerr Plant l!n i t 3 Reload 1, NEDO-24128 (Appendix A), June 1978 1
i n
509 192 i
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NEDO-24199 02 06 10 14 18 22 26 30 59 55 42 36 51 8
2 14 47 24 20 43 14 4
16 0
39 30 26 35 12 6
4 10 31 36 24 28 NOTES:
1.
ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC UPPER LEFT QUADRANT SHOWN ON MAP.
2.
NUMBERS INDICATE NUMBER OF NOTCHES WITHDRAWN OUT OF 48.
BLANK IS A WITHDRAbN ROD.
3.
ERROR ROD IS THE 30-43 ROD.
Figure 6.
Limiting RWE Rod Pattern 509 198
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NEDO-24199 1.2 LILTIMATE STABILITY LIMIT 1.0 08 N ATUR A L CIRCU LA TION
-o 5
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O o4 105% ROD LINE o2 o
o 20 40 eo 80 ioo 120 PERCENT POWER Figure 8.
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-24 C BOUNDING VALUE FOR 200 cal /g. COLD 0 BOUNDING VALUE FOR 280 ul/g, HS8 d CALCULATED VALUE - COLD Q CALCULATED VALUE - HSB l
l l
400 800 1200 1600 2000 2400 O
FUEL TEMPER ATURE ( C)
Figure 9.
Doppler Reactivity Coefficient Comparison for RDA 15 509 261
NEDO-24199 20 I
CALCULATED VALUE 16 E
a h
b 12 BOUNDING VALUE FOR 280 cal /g 5o Y'
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8 12 16 20 ROD POSITION, feet OUT Figure 10.
Accident Reactivity Shape Function at 20 C
O [
509
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NEDO-24199 20 s
'l 16 E,0NDING VALUE FOR 280 cal /g w
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CALCULATED VALUE N
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Figure 11.
Accident Reactivity Shape Function at 286 C 509 03 17
60 BOUNDING VALUE FOR 290 cal /g 50 G
N 9
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$c 20 10 0C C
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8 10 E LAPSED TIME (sec)
Figure 12.
Scram Reactivity Function at 20 C 509 204 18
N NEDO-24199 100 BOUNDING VALUE FOR 280 cal /g 80 G
N 9<
60 O
CALCULATED VALUE e
5 i
a C
40 5c 20 OC
~
0 2
4 6
8 10 ELAPSED TIME (sec)
Figure 13.
Scram React ivit: Function at 286 C "o
C c (), )
19/20 (j- (j r.)
NEDO-24199 APPFNDIX A Fue1 Loading Error LilGR:
16.02 kW/ft Safety /Relle Valve Capacity at Setpoint (No./%):
11/70 Spring Safety Valve Capacity at Setpoint (No./%):
2/14.2 ca., o
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