ML20083F215
| ML20083F215 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/30/1983 |
| From: | Charnley J, Hill R, Zarbis W GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20083F193 | List: |
| References | |
| Y1003J01A56, Y1003J01A56-R01, Y1003J1A56, Y1003J1A56-R1, NUDOCS 8312300200 | |
| Download: ML20083F215 (32) | |
Text
.....
Y1003J01A56 REVISION 1 CLASSI NOVEMBER 1983 1
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SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT RELOAD 5
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Y1003J01A56 Revision 1 Class I November 1983 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT RELOAD 5 Prepared:
W. A.
bis Verified.
s C. Hill Approved:
J/S. Charnley rael Licensing Mana er NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC i
i Y1003J01A56 Rev. 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared 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.
4 Nuclear Regulatory Commission (USNRC) for amending The Authority's operating license of the James A. FitzPatrick Nuclear Power Plant. The information cos-
- tained 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 June 12, 1970, 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 unauthorizcd 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 from such responsibility for liability or damage of any kind which may result use of such information.
11
-_ _ __ ______ _--_.~,___,_-_
?
Rev. 1 1.
PLANT UNIQUE ITEMS (1. 0)
- Appendix A: - GETAB Analysis Initial Conditions Appendix B: Verification of Operating Flexibility Options Appendix C: Full Arc Operation 2.
RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 and 4.0)
Fuel Type Cycle Loaded Number Number Drilled 1rradiated 8DRB283 3
12 12 P8DRB265L 4
24 24 P8DRB283 4-136 136 P8DRB284H 5
128 128 P8DRB299 5
60 60 New P8DRB299 6
200 200 Total 560 560 3.
REFERENCE CORE LOADING PATTERN (3.3.1) l Nominal previous cyc1e core average exposure at end of cycle:
17232 mwd /st
~
Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:
17050 mwd /st
[
Assumed reload cycle core average exposure at end of cycle:
17509 mwd /st j
Core loading pattern:
Figure 1 I
1
- ( ) Refers to area of discussion in " General Electric Standard Application L
for Reactor Fuel," NEDE-240ll-P-A-4, January 1982; a letter "S" preceding the number refers to the United States supplement.
l t
1
Y1003J01A56 Rev. 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)
Beginning of Cycle K-effective l
Uncontrolled 1.112 Fully Controlled 0.957 Strongest Control Rod Out 0.989 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, AK 0.000 5.
STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) ppm (20*C, Xenon Free) 600 0.026 6.
RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 and S.2.2)
(REDY EVENTS ONLY) r0C-1000 EOC-2000 EOC mwd /st mwd /st Void Fraction (%)
41.7 41.7 41.7 Average Fuel Temperature (*F) 1271 1271 1271 Void Coefficient
.N/A* (c/% Rg)
-9.19/-11.49
-9.99/-12.49
-10.37/-12.96 Doppler Coefficient N/A (c/*F)
-0.234/-0.222
-0.229/-0.218
-0.224/-0.213 Scram Worth N/A (S)
- 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-240ll-P-A-4, January 1982.
2
Y1003J01A56 Rev. 1 7.
RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Fuel-Peaking Factors R-Bundle Bundle Flow Initial Design Local Radial Axial Factor Power (MWt)
(1000 lb/hr)
MCPR Exposure: EOC 8x8R 1.20 1.45 1.40 1.051 6.193 116.8 1.31 P8x8R 1.20 1.42 1.40 1.051 6.059 117.7 1.34 Exposure: E0L-1000 mwd /st 8x8R-1.20 s'
1.40 1.051 6.338 116.0-1.28 P8x8R' 1.20 45 1.40 1.051 6.194 116.9 1.31 Exposure: EOC-2000 mwd /st 8x8R 1.20 1.54 1.40 1.051 6.550 114.8 1.24 P8x8R 1.20 1.51 1.40 1.051 6.424 115.6 1.26 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization: No Recirculation Pump Trip
- No Rod Withdrawal Limiter
- No Thermal Power Monitor
- Yes o
ODYN Option B Improved Scram Time
- Yes Exposure Dependent Limits. Yes
?
Exposure Points Analyzed : E0C, EOC-1000 mwd /st, E0C-2000 mwd /st 9.
CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
Exposure Flux Q/A ACPR Transient (mwd /st),
(% NBR)
(% NBR) 8x3R P8x8R Figure Load Rejection Without EOC 615 128 0.24 0.27 2a Bypass EOC-1000 550 125 0.21 0.24 2b E0C-2000 484 121 0.17 0.19 2c Loss of 80*F EOC 122 121 0.13 0.13 3
Feedwater Heating Feedwater Controller EOC 442 126 0.21 0.23 4a Failure E0C-1000 309 122 0.17 0.19 4b E0C-2000 248 118 0.13 0.14 4c 3
l l
J1003J01A56-Rev. 1 10.
LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(S.2.2.1)
Limiting Rod Pattern: Figure.5 I
Includes 2.2% Power Spiking -Penalty: Yes Rod Position Rod Block (ft fACPR MLHGR (kW/ft) 4 Reading withdrawn) 8x8R/P8x8R 8x8R/P8x8R 104-3.5 0.12 14.17 105 4.0 0.14 14.72 106 4.5 0.15 14.84 107 5.0 0.17 14.84 108 5.5 0.18 14.84 109 6.0 0.20 14.84 110 9.0 0.26 16.23 Set Point Selected:
108 11.
CYCLE MCPR VALUES (S.2.2)
Non-Pressurization Events Exposure Range: BOC to EOC P8x8R 8x8R Loss of 80*F Feedwater Heating 1.20 1.20 i
Fuel Loading Error 1.20 i-Rod Withdrawal Error 1.25 1.25 i
i -
l i
4 i
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Rev. 1 Y1003J01A56 Pressurization Events Option A Option B P8x8R 8x8R P8x8R 8x8R Exposure Raage:
BOC to EOC-2000 mwd /st Load Rejection Without Bypass 1.32 1.29 1.11 1.10 Feedwater Controller Failure 1.26 1.25 1.20 1.19 Exposure Range:
EOC-2000 mwd /st to EOC-1000 mwd /st Load Rejection Without Bypass 1.37 1.34 1.13 1.13 Feedwater Controller Failure 1.32 1.29 1.25 1.23 Exposure Range:
EOC-1000 mwd /st to EOC Load Rejection Without Bypass 1.40 1.37 1.28 1.25 Feedwater Controller Failure 1.36 1.34 1.29 1.27 12.
OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3) si v
Plant Transient (psig)
(psig)
Response
MSIV Closure (Flux Scram) 1218 1256 Figure 6 13.
STABILITY ANALYSIS RESULT (S.2.4)
Rod Line Analyzed: Extrapolated Rod Block Decay Ratio:
Figure 7 Reactor Core Stability Decay Ratio, x2 *0 0.93
/
Channel Hydrodynamic Performance Decay Ratio, x f*0 2
Channel Type 0.30 f
8x8R/P8x8R 5
Y1003J01A56 Rev. I 14.
LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial CPR Resulting CPR I
Misoriented 1.18 1.07
- 15.. 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:
Accident Reactivity Resultant Peak Enthalpy, Cold:
.225.5 cal /gm Parameter (s) not Bounded,' HSB :
Accident Reactivity Resultant Peak Enthalpy, HSB :
272.2 cal /gm
~ 16.
LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2) 4 See " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)," July 1977, NED0-21662 (as amended).
6
Rev. 1 Y1003J01A56 oMMMMMo sMMMMMMMMMs
- MEMMMMMMMMM
- eMMMMMMMMMMMo CMMMMMMMMMMMMM EMMMMMMMMMMMMM CMMMMMMMMMMMMM EMMMMMMMMMMMMM IMMMMMMMMMMMMM
- "MMMMMMMMMMM l'
MMMMMMMMMMM "MMMMMMMMM*
MMMMM" IIIIiIIIII, i
1 3 'i 7 9111315171921232527293133353739414345474951 n
FUEL TYPE A=
8DRB283 D=
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=
B
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=
Figure 1.
Reference Core Loading Pattern 7
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{
Rev. I Y1003J01A56 2
6 10 14 18 22 26 30 34 38 42 46 50 51 36 36 36 47 6
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NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48.
BLANK IS A WITHDRAWN ROD 2.
ERROR ROD IS (22,31).
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Reactor Core Decay Ratio versus Power 17
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Accident Reactivity Shape Function, Cold Startup 19
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Accident Reactivity Shape Function, llot Standby 20
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Scram Reactivity Function, Hot Standby 22
_ _ _ = _ _ _ _ - _ _ _ _ _ _ _
Y1003J01A56 Rev. 1 APPENDIX A GETAB ANALYSIS INITIAL CONDITIONS The values listed below were used in the GETAB analysis for this reload rather than the values given in Reference A-1, to more nearly reflect actual plant data.
Reactor Pressure (psia) 1035 Inlet Enthalpy (Btu /lb) 527.0 REFERENCE A-1.
" General Electric Standard Application for Reactor Fuel,"
NEDE-24011-P-A-4, January 1982.
A-1/A-2
Y1003J01A56 Rev. 1 APPENDIX B VERIFICATION OF OPERATING FLEXIBILITY OPTIONS The following operating flexibility options have been developed for BWRs.
A "Yes" indicates that the option has been verified as being applicable to Cycle 6.
1.
Single Loop Operation: Yes 2.
Load Line Limit: Yes 3.
Extended Load Line Limit: No 4.
Increased Core Flow: No 5.
Feedwater Temperature Reduction: No B-1/B-2
Rev. 1 Y1003J01A56 APPENDIX C FULL ARC OPERATION The analysis results summarized in the body of this document were performed assuming a partial are turbine control valve configuration. The change from a partial arc configuration to a full are configuration impacts only the Load Rejection Without Bypass event, since for this event it is the closure of the turbine control valves that causes a reduction in steam flow that results in a nuclear system pressure increase. For the Turbine Trip Without Bypass and the Feedwater Controller Failure events, the steam pressure increases as the result of closure of the turbine stop valves, so these events are not impacted by the change in turbine control valve configuration. Non-pressurization events are also not affected.
This appendix summarizes the results of the reanalysis of the Load Rejection Without Bypass event assuming full arc turbine control valve configuration.
The reload unique GETAB transient analysis initial condition parameters are listed in Table C-1.
.The core-wide transient analysis results are shown in Table C-2, and the cycle MCPR values are provided in Table C-3.
Although other events are not affected by the change in turbine control valve configuration, they are included in Table C-3 in order to determine cperating limits.
The results of this appendix supercede the results reported in the body of this submittal if a full are turbine control valve configuration is used.
C-1
Y1003J01 A56 Rev. 1 Table C-1 l
RELOAD UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS
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a ing actors Fuel R-ow r Bundle Flow Initial Design Local Radial, Axial Factor (MWt)
(1000 lb/hr)
MCPR Exposure. EOC 8X8R 1.20 1.44 1.40 1.051 6.148 117.1 1.32 P8X8R 1.20 1.41 1.40 1.051 5.990 118.1 1.36 Exposure: EOC - 1000 mwd /st 8X8R 1.20,
1.48 1.40 1.051 6.304 116.2 1.29 l
P8X8R 1.20 1.45 1.40 1.051 6.178 117.0 1.32 Table C-2 CORE WIDE TRANSIENT ANALYSIS RESULTS Expcsure Flux Q/A Transient (mwd /st)
(% NBR)
(% NBR) 8X8R P8X8R Figure Load Rejection EOC 678 128 0.25 0.29 C-1 Without Bypass EOC-1000 590 125 0.22 0.24 C-2 s
C-2
Rev. 1 Y1003J01A56 Table C-3 CYCLE MCPR VALUES Non-Pressurization Events Exposure Range:
BOC te EOC P8X8R 8X8R I,oss of 80*F Feedwater licating 1.20 1.20 1.20 Fuel Loading Error 1.25 1.25 Rod Withdrawal Error i
Pressurization Events Option A Option B P8X8R 8X8R P8X8R 8X8R Exposure Range:
BOC to EOC 1000 mwd /st Load Rejection Without Bypass 1.37 1.35 1.15 1.14 Feedwater Controller Failure 1.32 1.29 1.25 1.23 Exposure Range:
EOC-1000 mwd /st to EOC 1.42 1.38 1.20 1.26 Load Rejection Without Bypass Feedwater Controller Failure 1.36 1.34 1.29 1.27 C-3
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