ML20041A997

Supplemental Reload Licensing Submittal for James a Fitzpatrick Nuclear Power Plant Reload 4.

ML20041A997
Person / Time
Site:	FitzPatrick
Issue date:	02/28/1982
From:	Charnley J, Engel R, Leaser J GENERAL ELECTRIC CO.
To:
Shared Package
ML20041A993	List: ML20041A992 ML20041A997
References
Y1003J01A25, Y1003J01A25-R01, Y1003J1A25, Y1003J1A25-R1, NUDOCS 8202230135
Download: ML20041A997 (28)

Text

- .

Y1003J01A25 Revision 1 Class I February 1982 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT RELOAD 4 Prepared: 08b J'. D. L' easer l

Verified: --

[S. Charnley' J

Approved:

R. E. Engef, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORN! A 95125 GENER AL h ELECTRIC 8202230135 820219 PDR ADOCK 05000333 i P PDR 1

d Y1003J01A25 R:,v. 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This repcrt 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.

Nuclear Regulatory Commission (USNRC) for amending The Authority's operating license of the James A. FitzPatrick Nuclear Power Plant. The information con-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 inforwoiion 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.

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Y1003J01A25 Rev. I

1. PLANT UNIQUE ITEMS (1.0)*

Transient Analysis Initial Conditions: Appendix A Load Line Limit Analysis Reverification: Appendix B

2. RELOAD FUEL BUNDLES (1.0, 2.7, 3.3.1 and 4.0)

Fuel Cycle Designation Loaded Number Number Drilled Irradiated 8DB274L 2 20 20 8DB274H 2 56 56 8DRB265L 3 36 36 8DRB283 3 100 100 P8DRB265L 4 24 24 P8DRB283 4 136 136 New P8DRB284H 5 128 128 P8DRB299 5 60 60 Total 560 560

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 16.7 GWd/T l

Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 16.4 GWd/T Assumed reload cycle core average exposure at end of cycle: 17.2 GWd/T Core loading pattern: Figure 1

• ( ) refers to areas of discussion in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-240ll-P-A-1 and NEDO-240ll-A-1, July 1979, as revised by amendments 2-10.

1

Y1003J01A25 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)

BOC k gg Uncontrolled 1.116 Fully Controlled 0.957 Strongest Control Rod Out 0.987 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)

Iggi (20'C, Xenon Free) 600 0.03

6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)(1)

EOC-1 EOC GWd/T Void Coefficient N/A* (c/% Rg) -8.8/-11.0 -9.6/-12.0 Void Fraction (%) 41.7 41.7 Doppler Coefficient N/A (c/*F) -0.23/-0.22 -0.23/-0.22 Average Fuel Temperature (*F) 1243 1343 Scram Worth N/A (%)( }

~

Scram Reactivity vs Time (

• N = Nuclear Input Data A = Used in Transient Analysis (1) Applies to Inadvertent Startup of HPCI Pump Event Only (2) Generic, exposure independent values are used as given in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-240ll-P-A-1, Amendment 10, April 1981.

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Y1003J01A25 Rev. 1

7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

Peaking Factors Bundle Flow Fuel Exposure (Local, Radial, Bundle Power Initial-3 Design (GWd/T) Axial) R-Factor (MWt) (10 lb/hr) MCPR 8x8 EOC 1.22, 1.35, 1.40 1.10 5.75 115 1.29 E0C-1 1.22, 1.39, 1.40 1.10 5.92 114 1.25 8x8R EOC 1.20, 1.50, 1.40 1.05 6.40 115 1.29 EOC-1 1.20, 1.54, 1.40 1.05 6.58 114 1.25 P8x8R EOC 1.20, 1.48, 1.40 1.05 6.29 116 1.31 EOC-1 1.20, 1.52, 1.40 1.05 6.46 115 1.28

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

Transient Recategorization: No

~

Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes Measured Scram Time: No Exposure Dependent Limits: Yes Exposures Analyzed (GWd/T): EOC EOC-1

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Exposure Range & Q/A ACPR Transient (GWd/T) (% NBR) (%) 8x8/8x8R P8x8R Figure Load Rejection EOC 653 125 0.22 0.24 3a without Bypass EOC-1 609 122 0.18 0.21 3b Inadvertent Start BOC to 128 120 0.14 0.15 4 OC of HPC1 Pump Feedwater EOC 362 122 0.17 0.19 Sa Controller Failure EOC-1 311 120 0.15 0.16 Sb 3

$t .Y1003J01A25 - -

Rev. 1 J

10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(5.2.1) 4

, Limiting Rod Pattern: Fifure 6-Includes 2.2% Power Spiking Penalty: Yes Rod Position

Rod Block (Feet- ACPR- MLHGR (kW/ft)'

EReading* Withdrawn) 8x8**8x8R/P8x8R 8x8**8x8R/P8x8R

104 3.0 0.12. 16.1

) 105 3.5 0.15 16.5

. 106 4.0 -0.17 16.8 107 5.0 0.20- 16.9 108* 6.5 0.24 16.9 109 -7.0 0.24. 16.9' 110 8.0 -0.25 16.9

11. CYCLE MCPR VALUES (5.2)

.i j Exposure Range (GWd/t) Pressurization Events Option A ' Option B BOC to EOC-1 8x8/8x8R/P8x8R 8x8/8x8R/P8x8R t

- Load Rejection w/o Bypass 0.23/0.23/0.27 0.04/0.04/0.06 1

Feedwater Controller Failure 0.19/0.20/0.21 . O.13/0.14/0.15 j EOC-1 to EOC i ,

i Load Rejection w/o Bypass- 0.28/0.28/0.30 0.16/0.16/0.18 Feedwater Controller Failure 0.22/0.22/0.25 0.16/0.16/0.18' BOC to EOC Non-Pressurization Events 8xd/8x8R/P8x8R s

i Inadvertent HPCI Pump Start -0.14/0.14/0.15

-Rotated Bundle Error -/-/0.13 Rod Withdrawal Error -- 0.24/0.24

• Indicates set point selected.

j. **Not Limiting i

! 4

--- -~nl-- > *-S* - +-- +-wr* -m

E I-Y1003J01A25 Rev. I

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3) si v Plant Transient' (psig)~ (psig) Response MSIV Closure 1236 1275 Figure 7 (Flux Scram)

13. STABILITY ANALYSIS RESULTS (5.4)

Rod Line Analyzed: Extrapolated Rod Block Decay' Ratio: Figure 8 Reactor Core Stability Decay Ratio, x2 /*0: 0.87 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 8x8 Channel: 0.37 8x8R/P8x8R Channel: 0.30 .

14. ROTATED BUNDLE ERROR RESULTS (5.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Initial Resulting Resulting MCPR MCPR T.HGR (kW/ft) 1.18 1.07 15.32

15. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Bounding Analysis Results:

Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and' 13

16. LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (5.5.2)

See " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear. Power Plant (Lead Plant)," July 1977, NEDO-21662, as amended.

5

i Y1003J01A25' Rev. 1

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Y1003J01A25 Rev. 1 DELETED See Section 6 i

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-Y1003J01A25 Rev. 1 i

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Upper Left Quadrant'Shown on Map.

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l_ 3. Error Rod Is. Rod (22,31).

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12

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l Figure 8. Reactor Cdre Decay Ratio versus Power 14

Y1003J01A25 Rev. 1

0. O A CAL CULATED VA .UE-COLD B CAL CULATED VA .UE-HSB C BOWD VAL 280 CAL /G COLO

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Y1003J01A25 Rev. 1 0.020 A ACCIDENT FUNCTION B BDUNDING VALUE 280 CAL /G 0.015 gC C 0 M

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l Y1003J01A25 Rev. I

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17

Y1003J01A25 Rev. 1 .

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Y1003J01A25 Rev. 1 0.06 A SCF AM FUNCTIO 4 8 BOL NJING VALU : 280 CAL /G 0.05  !

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Y1003J01A25 Rev. 1 APPENDIX A TRANSIENT ANALY3IS INITIAL CONDITIONS S/RV Capacity 84.2%

This valve more accurately represents the capacity of the S/RVs in the "as-installed" conditions.

, 21/22

Y1003J01A25 Rav. 1 APPENDIX B LOAD LINE LIMIT ANALYSIS REVERIFICATION B.1

SUMMARY

The Load Line Limit Analysis (LLLA) previously performed for James A. FitzPatrick Nuclear Power Plant has been reverified for applicability to Cycle 5. It is concluded that analyses support operation above the 100% power / flow line within the region defined in Reference B-1 for Cycle 5.

B.2 INTRODUCTION The purpose of this supplement is to provide documentation which supports opera-tion above the 100% power / flow line for James A. FitzPatrick Nuclear Power Plant for Reload 4, Cycle 5. The region of operation supported by this supple-ment is the same as the region defined and supported in Reference B-1.

B.3 BACKGROUND An analysis was previously performed which supported operation above the 100%

power / flow line and was documented in Reference B-1.

The generic applicability of the LLLA to BWR/4 plants for follow-on cycles-has been reviewed. It was concluded that only the standard reload analyses are required for licensing purposes to justify operation in the extended region identified in Reference B-1 for all cycles, with three special considerations as follows:

-1. Stability - A stability analysis shall be performed of the extended APRM rod / block line power / natural circulation flow.

2. ECCS 'the ECCS analyses previously submitted to the NRC shall be veri-fied on a cycle-by-cycle basis.

3. Transients - If the feedwater controller failure event is limiting, special analyses must be performed to determine if operating limit adjustments are necessary for operation in the extended operating range.

. 23

aussewnsww m )

~. ;

If th2 originsi LLLA was performed on a REDY trhnsient cnnlysic b sis, the limiting pressurization event will be evaluated at the 100% power, reduced flow condition with ODYN to verify that the licensing basis point (104,100) is still the most limiting point.

All other analyses (including the transient analyses) performed for each reload were determined to be bounding for operation in the extended region.

B.4 STABILITY ANALYSIS A stability analysis has been performed at the extended APRM rod block line power / natural circulation flow point with acceptable results as reported in Reference B-2.

B.5 ECCS ANALYSIS Both the previous ECCS analyses for FitzPatrick and the ECCS analyses for Cycle 5 have been verified for applicability of operation in the extended region defined by the Load Line Limit Analysis.

B.6 TRANSIENTS Both the Load Rejection without Bypass and the Feedwater Controller Failure events were analyzed with the ODYN code at the 100% power, 94% flow point to verify that the licensing basis point is still the most limiting for both the change to the ODYN transient analysis code and for the Cycle 5 core loading.

Results of these analyses and the licensing basis analyses are given in Table B-1. As shown, the licensing basis values remain the most limiting.

B.7 CONCLUSIONS The required cycle-specific and ODYN analyses for application of the Load Line Limit Analysis have been performed with acceptable results, and, therefore, operation in the extended region above the 100% power / flow line, as defined in Reference B-1, is acceptable for James A. FitzPatrick Nuclear Power Plant.

24

Y1003J01A25 Rev. 1

.B.8 RSFERENCES B-1. " General Electric Boiling Water Reactor Load Line Limit Analysis for James A. FitzPatrick Nuclear Power Plant," February 1980 (NEDO-24243).

B-2. " Supplemental Reload Licensing Submittal for James A. FitzPatrick Nuclear Power Plant Reload 4," August 1981 (Y1003J01A25).

.25

7 Y1003J01A25 , ,

Rev. 1 ~.

Table B TRANSIENT RESULTS FOR LLLA VERIFICATION Power Core Flow $ Q/A Transient Exposure (%) (%) (%-NBR) (%)

Load Rejection without Bypass EOC 104 100 653 125 Load Rejection without Bypass EOC 100 94 596 120 Feedwater Controller Failure EOC 104 100 362 122 Feedwater Controller Failure EOC 100 94 332 118 i

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, 26

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