ML20151M887
| ML20151M887 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/30/1988 |
| From: | Charnley J, Lambert P, Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19292J200 | List: |
| References | |
| 23A5898, 23A5898-R, 23A5898-R00, NUDOCS 8808080029 | |
| Download: ML20151M887 (24) | |
Text
i 23A5898 REVISION 0 CLASS I JUNE 1988 O
O (23A5898, REV. 0)
SUPPLEMENTAL RELOAD LICENSING REPORT O FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT RELOAD 8 (CYCLE 9)
O O
Prepared: .
P. A. Lambert Fuel Licensing O
Verified:
J.'L". Rash /
w-M i Fuel Licensing Approved: /
. Cha W yi tanager O el Licensing O
v 8008080029 880729 PDR ADOCK 05000333 P PDC GENuclearEnergy 115 Cunner Avenue A O SanJose. CA 951?S 1/2
l 23A5898 Rev. O O .
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY O
This report was prepared by General Electric Company (GE) solely for the Power Authority of the State of New York (The Authority) for The O 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 GE to be an accurate and true representation of the facts known, O obtained or provided to GE 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 O Authority and GE 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
'O 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 GE nor any of the contributors to this document or that such use of such information may not infringe privately owned rights; nor do they assume any O responsibility for liability rr damage of any kind which may result from such use of such information.
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ACKNOWLEDGEMENT g The erigineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Report were Performed by G. M. Baka and J.L. Casillas of the Fuel Engineering Section.
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- 1. PLANT-UNIQUE ITEMS (1.0)*
Plant Parameter Differences Appendix A C) Use of GEXL-PLUS Methods for Cycle 9 Appendix B Use of New Safety Limit MCPR for Cycle 9 Appendix C
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
O Fuel Type Cycle Loaded Number Irradiatrd BP8DRB299 7 188 C) BP8DRB299 8 4 BD319A 8 184 New C)
BD339A 9 32 BD336A 9 152 Total S60 0
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Naminal previous cycle core average exposure at
() end of cycle: 22003 mwd /MT Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 21673 mwd /MT Assumed reload cycle core average exposure at
() end of cycle: 22570 mwd /MT Core loading pattern: Figure 1 O
- ( ) Refers to area of discussion in "General Electric Standard
() Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. A lettor "S" preceding the number refers to the U.S. Supplement, NEDE-24011-P-A-8-US, May 1986.
C)
23A5898 Rr>v. 0 O
- 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 eff 9 Uncontrolled 1.117.
Fully Controlled 0.970 Strongest Control Rod Out 0.989 ,
R, Maximum Increase in Cold Core Reactivity 0.001 with Exposure into Cycle, delta k
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.}), g Shutdown Margin (delta K) ppm (20*C, Xenon Free) 600 0.030 9
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(Cold Water Injection Events Only)
Void Fraction (%) 42.2 g Average Fuel Temperature (*F) 1077 Void Coefficient N/A* (6/% Rg) -7.96/-9.95 Doppler Coefficient N/A* (6/*F) -0.238/-0.226 g Scram Worth N/A* ($) **
O 9
- N = Nuclear Input Data, A = Used in Transient Analysin
- Generic exposure independent values are used as given in "General 9 I Electric Standard Application for Reactor Fuel." NEDE-24011-P-A-8, l dated May 1986.
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- 7. RELOAD-UNIQUE GETAB_ TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) /
Bundle Fuel Peaking Factors Power Bundle Flow Initial
() Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) (MCPR)
Exposure: EOC9 BP8x8R 1.20 1.48 1.40 1.051 6.286 112.0 1.28
() GE8X8EB 1.20 1,48 1.40 1.051 6.294 114.4 1.29 Exposure: E0C9-1102 mwd /MT BP8x8R 1.20 1.53 1.40 1.051 6.499 110.8 1.24
() GE8x8EB 1.20 1.54 1.40 1.051 6.517 113.1 1.24 Exposure: E0C9-2205 mwd /MT BP8x8R 1.20 1.56 1.40 1.051 6.627 110.0 1.21
() GE8x8EB 1.20 1.56 1.40 1.051 6.610 112.5 1.22
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
O Transient Recategorization No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes Improved Scram Time: Yes (ODIN Option B)
() Exposure Dependent Limits: Yes Exposure Points Analyzed: 3
- 9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)
O Single-Loop Operation: Yes Load Line Limit: Yes Extended Load Line Limit: No Increased Core Flow: No C) Flow Point Analyzed: N/A Feedwater Temperature Reduction: No ARTS Program: No Maximum Extended Operation Domain: No C)
()
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- 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
Methods Used: GFMINI 9
Flux Q/A Delta CPR Transient (% NBR) (% NBR) BP8x8R GE8x8EB Figure Exposure Range: BOC9 to EOC9 O
Loss of Feedwater Heating 119 117 0.12 0.12 2 Exposure Range: E0C9 Load Rejection Without Bypass 653 128 0.24 0.25 3 g Feedwater Controller Failuro 454 124 0.20 0.20 4 Exposure Range: E0C9-1102 mwd /MT Load Rejection Without Bypass 485 123 0.20 0.20 5 g Feedwater Controller Failure 339 119 0.15 0.16 6 Exposure Range: E0C9-2205 mwd /MT Load Rejection Without Bypass 420 121 0.17 0.18 7 g Feedwater Controller Failure 300 116 0.13 0.13 8
- 11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(S.2.2.1)
O Limiting Rod Pattern: Figure 9 Rod Block Rod Position Delta CPR MLHGR (kW/ft)
Reading (%) (Feet Withdrawn) BP8x8R GE8x8EB BP8x8R GE8x8EB 104 3.0 0.11 0.11 14.50 15.83 e 105 3.5 0.14 0.14 15.04 16.54 106 4.0 0.16 0.16 15.59 17.25 107 4.5 0.18 0.18 15.78 17.64 108 6.0 0.24 0.24 16.37 18.80 109 12.0 0.29 0.29 16.87 20.05 110 12.0 0.29 0.29 16.87 20.05 9j l
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- 12. CYCLE MCPR VALUES (S.2.21 Non-Pressurization Events O Exposure Range: BOC to EOC BP8x8R GE8x8EB Loss of Feedwater Heating 1.16 1.16
(). 1.23 Fuel Loading Error --
Rod Withdrawal Error 1.22 1.22
() Pressurization Events
- Oytion A Option B BP8x8R GE8x8EB BP8x8R GE8x8EB Exposure Range: EOC9 O
Load Rejection Without Bypass 1.34 1.34 1.30 1.30 Feedwater Controller Failure 1.29 1.28 1.26 1.25 C) - Exposure Range: EOC9-1102 mwd /MT Load Rejection Without Bypass 1.34 1.34 1.27 1.27 Feedwater Controller Failure 1.25 1.26 1.23 1.24 C)
Exposure Range: E0C9-2205 mwd /MT Load Rejection Without Bypass 1.31 1.32 1.24 1.25 Feedwater Controller Failure 1.22 1.23 1.20 1.21 0
- 13. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3)
P P s1 -v
() Transient (psia) (psin) Plant Response MSIV Closure 1223 1261 Figure 10 (Flux Scram)
I C) *0DYN Adjustment Factors are documented in a letter from J S. Charnley (GE) to H.N. Berkow (NRC), "Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE-24011-P-A " Jaauary 16, 1986.
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- 14. LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Event Delta CPR O' Misoriented 0.19
- 15. CONTROL R0D DROP ANALYSIS RESULTS (S.2.5.1) g; Plant specific Analysis Results:
Parameter (s) not Bounded, Cold: None Resultant Peak Enthalpy, Cold: N/A gi Parameter (s) not Bounded, HSB: None Resultant Peak Enthalpy, HSB: N/A 9: '
- 16. STABILITY ANALYSIS RESULTS (S.2.4)
GE SIL-380 recommendations have been included in the James A.
Fitzpatrick Nuclear Power Plant operating procedures and/or Technical Specifications and, therefore, no stability analysis is required. g.
See "James A. Fitzpatrick Nuclear Power Plant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31317P, October 1986 (as e amended). j l
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- 17. LOSS-OF-COOLANT ACCIDENT RESULTS (S.2.5.2) (continued)
() Technical Specification MAPLHGR Limits Fuel Type: BD339A (GE8x8EB)*
MAPLHGR (KW/ft)
Average Planar Exposure
() (GWd/ST) Most Limiting Least Limiting 0.2 11.0 11.4 1.0 11.2 11.5 4.0 11.9 12.2 5.0 12.0 12.3
() 10.0 12.9 12.9 12.5 12.7 12.8 15.0 12.6 12.6 25.0 11.6 11.6 35.0 10.4 10.4 45.0 8.8 8.9 C) 50.0 6.2 6.3 Fuel Type: BD336A (GE8x8EB)*
%D MAPIEGR (KW/ft)
Average Planar Exposure (GWd/ST) Most Limiting Least Limiting 2 0.2 10.9 11.5 1.0 11.1 11.6 C) 2.0 11.2 11.7 -
3.0 11.4 11.9 7.0 12.0 12.5 10.0 12.6 12.8' 12.5 12.7 12.8 15.0 12.5 12.5 C) 25.0 11.4 11.5 35.0 10.3 10.3 45.0 8.5 8.7 50.0 6.0 6.2 O
- The GE8x8EB LOCA analysis results presented in Section 5 of NEDC-31317P, bound the MAPLHGR limits for the BD339A and BD336A fuel types since the C) analysis was performed at 14.0 KW/ft. This analysis yielded a licensing basis peak clad temperature of 1573 F and peak oxidation fraction of
<0.50%
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23A5898 Rnv. 0
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- "MMMMMMMMM" l "MMMMM" l l l l l l l l l l 1 3 57 9111315171921232527293133353739414345474951 O FUEL TYPE A = BD319A D = BP8DRB299 B = BD336\ E = BP8DRB299 C = BD339A 4
Figure 1. Reference Core Loading Pattern l
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k 23A5898 Rev. 0
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- . Figure 2. Plant Response to Loss of 80*F Feedwater i Heating (BOC9 to E0C9) iO
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23A5898 Rev. 0 .
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Figure 3. Plant Response to denerator Load Rejection Without Bypass (EOC9) g 16 91
23A5898 Rav. O O .
O no. e i NEurRON N X 1 YES !EL PRE .'91SE d SI) 2 ave S E HE T FLUX 2 SAF :TY VAL E Low 3 COR IP FLC 3 REL lEF VAL E LOW e --e 15 0. e
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23A5898 Rev. 0 ..,
!hEUIRON FLLx 1 YESSEL PRESb RISE (PSI) 2 AVE SURr ACE HE AT FLUX 2 Sargiy vntst rLcv 3 CORE INLET r t; .- 3 RELIEr V ALst FLOW 15 0. 0 300.0 ' P ei-- '* _
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23A5898 Rev. 0 40 .
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Figure 6. Plant Res,'onse to Feedwater Controller O Failure (E009-1102 mwd /MT) 19 O
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23A5898 Rev. 0 . ,m l
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1 NEUTRDN FL M 1 VESSEL PRES 5 RISE (PSI) 2 AVE SLSTA E HEAT FLUX 2 SAFETY val \E FLCv 3 CCRE INLET TLOW 3 RELIEF Y AL'.E FLod 1
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. 23A5898 Rev. 0
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Failure (EOC9-2205 mwd /MT)
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23A5898 Rev. 0 *
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- l 2 6 10 14 18 22 26 30 34 38 42 46 50 51 36 36 36 47 6 6 6 6 g
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l NOTES: gl
- 1. No, indicates number.of notches withdrawn cut of 48. Blank is l a Withdrawn Rod.
- 2. Error rod is (22, 31).
O Figure 9. Limiting Rod Pattern 9
I 22 1 9i
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! 1 NEUTRON i UX 1 YESSEL PS $5 t!SE(P$1) 2 AVE SURr E HE AT FLUX 2 SAFETY VA VE FL0s
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Figure 10. Plant Response to MSIV Closure, Flux scram l
23/24 0
23A5898 Rsv. 0 0 .
APPENDIX A PLANT PARAMETER DIFFERENCES O
GETAB and Transient Analysis Initial Conditions The values used in the GETAB and Transient Analysis which differ from !
O the values reported in Tables S.2-4.1, S.2-4.2, S.2-6 and S.2-8 in NEDE-24011-P-A-8-US are given in Table A-1.
Table A-1 O
PLANT PARAMETER DIFFERNCES Parameter Analysis Value NEDE-24011-P-A-8-US Value Q Thermal Power (MWt)1 2436 2535 1 0.2%
Rated Steam Flow (1b/hr)1 10.47E+06. 10.96E+06 1 0.2%
Dome Pressure (psi)1 1006 1020 1 2 psi
'O- Turbine Pressure (psi)1 956 960 1 2 psi Non-Fuel Power Fraction 1 0.039 0.040 O
1 The indicated changes are a result of the applicatien of the pre-approved methods outlined in Amendment 11 to NEDE-24011-P-A-8.
O O
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23A5898 Rsv. C APPENDIX B USE OF'GEXL-PLUS CORRELATION FOR CYCLE 9
()
The analyses required for this cycle were performed with the GEXL-PLUS
() thermal correlation. In analyses prior to Cycle 9 (Reload 8), the GEXL thermal correlation was used. The incorporation of GEXL-PLUS into the fuel cycle analysis process is provided for in Amendment 15 to GESTAR-II i (NEDE-24011-P-A-8).
C) 0 0
0 0
0 0
27/28
()
.3A5898 R:v. O iO APPENDIX C USE OF NEW SAFETY LIMIT MCPR FOR CYCLE 9 O
The analyses required for this cycle were performed with the upgraded O safety limit MCPR of 1.04, instead of the previous safety limit MCPR of 1.07.
The implementation of this safety limit is a result of the utilization of fuel types with high bundle P-factors, as stipulated in Amendment 14 to GESTAR-II (NEDE-24011-P-A-8).
O O
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s 29/30 (FINAL)
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