ML20096H131

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Rev 0 to Supplemental Reload Licensing Submittal for Peach Atomic Power Station Unit 2,Reload G
ML20096H131
Person / Time
Site: Peach Bottom Constellation icon.png
Issue date: 06/30/1984
From: Elliott P
GENERAL ELECTRIC CO.
To:
Shared Package
ML20096H087 List:
References
22A8597, 22A8597-R, 22A8597-R00, NUDOCS 8409110269
Download: ML20096H131 (30)
v • d • e


Text

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SUPPLEMENTAL RELOAD 4

LICENSING SUBMITTAL FOR ]
PEACH BOTTOM ATOMIC POWER STATION N UNIT 2, RELOAD 6 2 y

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22A8597 Revision 0 Class I June 1984 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 2, RELOAD 6

//

Prepared: -

  • P. E. Elliott Fuel Licensing Verified: )

R./f. Hill Fuel Licensing

, l Approve h ,-

. Charnley, Ma el Licensing NUCLEAR ENERGY BUSINESS OPERATIONS

  • GENERAL ELECTRIC COMPANY SAN JOSE CALIFORNIA 95125 GENERAL $ ELECTRIC 1/2

22A8597 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Philadelphia Electric Company (PECo) for PECo's use with the U. S. Nuclear Regula-tory Commission (USNRC) for amending PECo's operating license of the Peach Bottom Atomic Power Station Unit 2. The information contained 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 ac 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 Philadelphia Electric Company and General Electric for fuel bundles and services for reload fuel supply for Peach Bottom Atomic Power Station Units 2 and 3, dated November 16, 1979, and nothing con-tained 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 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.

3/4

23A8597 Rev. 0

1. PLANT UNIQUE ITEM (1.0)*

Appendix A - BP8DRB299H Bundle Appendix B - Safety Valve Capacity Appendix C - Single-loop Operation l

2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number Number Drilled Irradiated (P8x8R Fuel)

P8DRB285 5 40 40 P8DRB284H 5 40 40 P8DRB284H 5 68 68 P8DRB284H 5 48 48 l P8DRB299 6 124 124 P8DRB200 6 16 16 P8DRB284H 6 136 136 New (BP8x8R Fuel)

BP8DRB299 7 136 136 BP8DRB299H 7 156 156 Total 764 764 l 3. REFERENCE CORE LOADING PATTERN (3.3.1) l Nominal previous cycle core average exposure at end of cycle: 18460 mwd /ST l Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 18460 mwd /ST Assumed reload cycle core average exposure at end of cycle: 17941 mwd /ST Core loading pattern: Figure 1

  • ( ) kefers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-6, dated April 1983. A letter "S" preceding the number refers to the appropriate country-specific supplement.

5

22A3597 Rsv. 0

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,ff Uncontrolled 1.117 Fully controlled 0.962 Strongest Control Rod Out 0.987 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, AK 0.003

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (AK) ppm (20*C, Xenon Free) 660 0.035

6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)

(Cold Water Injection Events Only)

EOC7-2000 mwd /ST EOC7 Void Fraction (%) 39.8 39.8 Average Fuel Temperature (*F) 1279 1279 Void Coefficient N/A* (c/% Rg) -9.99/-12.49 -8.35/-10.44 Doppler Coefficient N/A (c/*F) -0.215/-0.204 -0.222/-0.211 Scram Worth N/A ($) **

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  • 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-6, dated April 1983.

6

22A8597 Rev. 0

7. RELOAD UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) ea ing Factors f Fuel ow Bundle Flow Initial l

Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR f

BOC 7 to EOC 7-2000 mwd /ST P8x8R 1.20 1.54 1.40 1.051 6.493 109.2 1.25 BP8x8R 1.20 1.54 1.40 1.051 6.498 109.2 1.25 EOC 7-2000 mwd /ST to EOC 7 P8x8R 1.20 1.45 1.40 1.051 6.123 111.6 1.33 BP8x8R 1.20 1.45 1.40 1.051 6.123 111.6 1.33

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: No Improved Scram Time: Yes Exposure Dependent Limits: Yes Exposure Points Analyzed: 2 l

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single-Loop Operation: Yes Load Line Limit: No Extended Load Line Limit: No Increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No 7

22A5987 Rev. 0

10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Flux Q/A Transient (% NBR) (% NBR) P8x8R BP8x8R Figure Exposure: BOC 7 to E0C 7-2000 mwd /ST Load Rejection w/o Bypass 548 122 0.18 0.18 2 Exposure: EOC 7-2000 mwd /ST to EOC 7 Load Rejection w/o Bypass 675 128 0.26 0.26 3 Exposure: BOC to EOC Loss of Feedwater Heater 125 123 0.15 0.15 4 Exposure: BOC 7 to EOC 7-2000 MRd/ST Feedwater Controller Failure 236 116 0.09 0.09 5 Exposure: EOC 7-2000 mwd /ST to E0C 7 Feedwater Controller Failure 321 124 0.18 0.18 6

11. IDCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRii4ENT FAILURE)

TRANSIENT SIEMARY (S.2.2.1)

Limiting Rod Pattern: Figure 7 Includes 2.2% Power Spiking Penalty: Yes Rod Position 0

Rod Block (Feet MLHCR Reading Withdrawn) P8x8R BP8x8R (kW/ft) 104 3.5 0.11 0.11 17.10 105 4.0 0.13 0.13 17.57 106 4.5 0.14 0.14 17.76 107 5.0 0.16 0.16 17.76 108 6.0 0.18 0.18 17.76 109 7.0 0.20 0.20 17.76 110 12.0 0.24 0.24 17.76 Setpoint Selected: 107 8

l h , _ , , _ _ _ _ _ _ ,

22A5987

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events _

Exposure Range: BOC 7 to EOC 7 P8x8R BP8x8R Loss of Feedwater Heater 1.22 1.22 Fuel Loading Error 1.22 1.22 Rod Withdrawal Error 1.23 1.23 Pressurization Events Exposure Range: BOC 7 to EOC 7-2000 mwd /ST Option A Option B P8x8R BP8x8R P8x8R BP8:8R Load Rejection w/o Bypass 1.30 1.30 1.11 1.11 Feedwater Controller Failure 1.21 1.21 1.15 1.15 Pressurization Events Exposure Range: EOC 7-2000 mwd /ST to EOC 7 Option A Option B P8x8R BP8x8R P8x8R BP8x8R Load Rejection w/o Bypass 1.39 1.39 1.27 1.27 Feedwater Controller Failure 1.30 1.30 1.24 1.24

13. OVERPRESSURIZATION ANALYSIS SUK' ARY (S.2.3) sl v Transient fpsig) (psig) Plant Response MSlV Closure 1244 1275 Figure 8 (Flux Scram)
14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: 105%

Decay Ratio: Figure 9 Reactor Core Stability Decay Ratio, x2 X0 / 0.87 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 Channel Type P8x8R/SP8x8R 0.29 9

22A8597 Rev. 0

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes i -

Event Initial MCPR Resulting MCPR Misoriented 1.20 1.07

16. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

Bounding Analysis Results:

Doppler Reactivity Coefficient Figure 10 Accident Reactivity Shape Functions: Figures 11 and 12 Scram Reactivity Functions: Figures 13 and 14 Plant Specific Analysis Results:

Parameter (s) not Bounded, Cold: Scram Reactivity Resultant Peak Enthalpy, Cold: 182.9 Parameter (s) not Bounded, HSB: Accident Reactivity Scram Reactivity Resultant Peak Enthalpy, HSB: 241.3

17. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)

See " Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2," December 1977 NEDO-24081, as amended.

10

22A8597 Rev. 0

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":""8:9, E = BP8DRB299H Figure 1. Reference Core Loading Pattern 11

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22A8597 Rev. 0 1 NEUTRON FLUK 1 VESSEL PRESS RISE (PSI) 2 AVE SURF ACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET LOW 3 RELIEF VALVE FLOW 150.0 300.0 ' evons? var _uE eLcy

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Figure 2. Plant Response to Load Rejection, Without Bypass, EOC 7-2000 mwd /ST 12

22A8597 Rev. 0 1 NEUTRON FLUK 1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET LOW 3 RELIEF VALVE FLOW 150.0 300.0 '.ovonss v>LtE rLew 100.0 N 200.0 N

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Figure 3. Plant Response to Load Rejection Without Bypass, EOC 7 13 I

=D 22A8597 Rev. 0 4

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-Figure 5. Plant Response to Feedwater Controller Failure, EOC 7-2000 mwd /ST j 1

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22A8597 Rav. 0 150.0 1 NEUTRON kir ( 1VESSELPRESlSRISE(PSI) 2 AVE SURF TE AT FLUX 2 SAFETY VA FLOW T TL W 3 RELIET VAL FLOW

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Figure 6. Plant Response to Feedwater Contrclier Failure. EOC 7 16

22A8597 Rsv. O n

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NOTES: 'l.. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC.

2. NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN
t. OUT OF 48. BLANK IS A WITHDRAWN ROD.
3. JERROR R0D IS'(30, 35).

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22A8597 Rzv. O I 1 NEUTRON F.UX 1 VESSEL FEESS RISE (PSI) 2 AVE SURF 8:E HEAT FLUX. 2 SAFETY V/LVE FL0t'

.. 3 CORE INLET FLOW 3 REL IEF V/'VE FLOW _

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0. 0 5.0 0.0 5.0 TIME (SECONOS) TIME (SECONOSI Figure 8. Plant Response to MSlV Closure (Flux Scram) 18

4 22A8597 R;v. O AF ATURAL C :RCULATIO 4 81 05 PERCENT R00 LI 4E-CL LTIMATE STABILITY LINE 1.-00 (:  ::

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-40.0 0.'0 .500.0 1000.0 1500.0 2000.0 2500.0 3000.O FUEL TEMPERATURE DEG C.

Figure 10. -Fuel Doppler Reactivity Coefficient in 1/A*C ,

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22A8597 R;v. 0 l2 0F0 :-

A ACCIDENT FUNCTION B BOUNDING VALUE 280 CAL /G 1-7. 5 11 5 . O m

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.22A8597- R2v. 0 7

-2. 0 . O A ACCIDENT FUNC TION 8 BOUNDING VALU E 280 CAL /G

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22A8597 R;v. 0 30.O A SCRAM FU NCTION B BOUNDING VALUE 280 CAL /G 25.0-M ca l-w 20.0 r.

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2.h8597 Rev. 0 50.0 ' '

A SCRAM FL NCTIO'N B BOUNDING VALUE 280 CAL /G 40.0 A m

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22A8597 R;v. . O APPENDIX A BP8DRB299H BUNDLE

' Pertinent information for the BP8DRB299H bundle is found in Reference A-1. Because of a typographical error, the bundle is identified as BP8DRB299A in that reference. The error will be corrected when Reference A-1 is issued.

. REFERENCE A-1. Letter, J.'S. Charnley (GE) to C. O. Thomas (NRC), " Proposed Administra-tive Amendment to GE Licensing Topical Report NEDE-24011-P-A", January 25, 1984.

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l.

25/26

.i ..

22A8597 Rev. O APPENDIX B SAFETY VALVE CAPACITY The safety valve capacity provided in Table S.2-4b of Reference B-1 is 934120 lb/hr at a reference pressure of 1240 psig. A value of 925700 lb/hr at a reference pressure of 1230 psig was used in the analyses to more nearly reflect actual plant parameters.

REFERENCE B-1. NEDE-24011-P-A-6-US, " General Electric Standard Application for Reactor

' Fuel (U.S. Supplement)", April 1983.

l 27/28 t

22A8597 R:v. O APPENDIX C SINGLE-LOOP OPERATION The single loop operation analysis, Reference C-1, has been verified for applicability to Cycle 7 and the results for Cycle 7 are doctamented in Reference C-2..

REFERENCE l

C-1. " Peach Bottom Atomic Power Station Units 2 and 3 Single Loop Operation,"

General Electric Company, May 1980 (NEDO-24229-1).

C-2. Erratta and Addenda No. 2 to NEDO-24229-1, dated June 1984.

l i

L 29/30 (FINAL)

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