ML20010H924
| ML20010H924 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/31/1981 |
| From: | Charnley J, Engel R, Leaser J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20010H915 | List: |
| References | |
| Y1003J01A28, Y1003J1A28, NUDOCS 8109290417 | |
| Download: ML20010H924 (19) | |
Text
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- T3sTl l
AUGUST 1981
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SUPPLEMENTAL RELOAD LICENSING pl SUBMITTAL FOR PILGRIM NUCLEAR
() POWER STATION UNIT 1, RELOAD 5 o
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O kONEoNEjjg GENER AL h ELECTRIC
I Y1003J01A28 Revision 0 Class I August 1981 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PILGRIM NUCLEAR POWER STATION UNIT 1, RELOAD 5 l
Prepared: O[ ,44/4~-
'J. D. Leaser Verifie '
I 6
/ J. S. Charnley f' Approved: A s7 R. E. Enge}f Manager Reload Fuel Licensing i
NUCLEAR POWER SYSTEMS OlVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC 1
Y1003J01A28 Riv. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Boston Edison Company (BECo) for BECo's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending BECO's operating license of the Pilgrim Nuclear Power Station. /
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or pro- ,
vided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting informa-tion in this dace 2nt are contained in the contract between Boston Edison Company and Ge...al Electric Company for reload fuel fabrication for the nuclear system for Pilgrim Nuclear Power Station, dated July 14, 1972, and nothing contained in this document shall be construed as changing said con-tract. 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 completeneas, accuracy or useful-ness 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 o f any kind whic: may result from such use of such information.
11
Y1003J01A28 sv. O r
- 1. PLANT UNIQUE ITEMS (1.0)*
New Control Rod Withdrawal Error Analysis Procedure: Appendix A
- 2. RELOAD FUEL BUNDLES (1.0, 2.7, 3.3.1 AND 4.0)
Fuel Cycle Designation Loaded Number Number Drilled Irradiated 8DB219H 4 68 68 8DB219L 4 156 156 P8DRB265L 5 120 120 P8DRB282 5 64 64 New P8DRB265H 6 60 60 P8DRB282 6 112 112 Total 580 580
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at 14.0 GWd/T end of cycle:
Minimum previous cycle core average exposure at 14.0 GWd/T end of cycle from cold shutdown considerations:
Assumed reload cycle core average exposure at 15.3 GWd/T end of cycle:
Core loading pattern: Figure 1
- ( ) refers to areas of discussion in " General Electric Boiling Water Reacter Generic Reload Fuel Application," NEDE-240ll-P-A-1 and NED0-240ll-A, July 1979.
'l i
i Y1003J01A28 R;v. 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)
BOC kggg Uncontrolled 1.113 Fully Controlled 0.952 Strongest Control Rod Out -
0.985 R, Maximum Increase in Cold Core Reactivity 0.001 with Exposure Into Cycle, Ak
- 5. STANDBY LIQUID CONTROL. SYSTEM SHUTDOWN CAPAEILITY (3.3.2.1.3)
Shutdown Margin (Ak) 3 (20*C, Xenon Free) 700 0.05 l 1
- 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)(1)
EOC6 Void Coefficient N/A* (C/* Rg) -6.1/-7.6 Void Fraction (%) 36.9
{
Doppler Coefficient N/A (C/ F) -0.22/-0.21
\verage Fuel Temperature (*F) 1205 Scram Worth N/A ($)(
Scram Reactivity vs Time ( }
- N = Nuclear Input Data A = Used in Transient Analysis 2 Applies to Loss of Feedwater Heating Event only.
Generic, exposure independent values are used as given in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-24011-P-A-1, Amendment 10, April 1981.
2
Y1003J01A28 Rev. 0
- 7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)
Peaking Factors "" " "
Fuel Exposure (Local, Radial, Bundle Power Initial 3
Design (GWd/T) Axial) R-Factor (MWt) (10 lb/hr) MCPR 8x8 EOC6 1.22, 1.61 1.10 5.41 98 1.31 1.40 P8x8R EOC6 1.20, 1.74 1.05 5.84 99 1.33 1.40
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: No Measured Scram Time: No Exposure Dependent Limits: No
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Exposure Range $ Q/A (GWd/T) (% NBR) (%) 8x8 P8x8R Figure Load Rejection EOC6 589 119 0.24 0.27 3 without Bypass Loss of 100*F EOC6 117 115 0.14 0.15 4 Feedwater heater Feedwater EOC6 313 116 0.19 0.20 5 Controller Failure
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(5.2.1)
See Appendix A.
3
Y1003J01A28 Ray, O
- 11. CYCLE MCPR VALUES (5.2)
Exposure Range (GWd/t) Pressurization Events Option A Option B BOC to EOC 8x8 _P8x8R 8x8 P8x8R Load Rejection w/o 0.30 0.33 0.25 0.28 Bypass Feedwater Control 1.er 0.25 0.26 0.17 0.18 Failure Nonpressurization. Events 8x8 P8x8R Loss or Feedwater Heating 0.14 0.15 Rotated Bundle Error --
0.17 Rod Withdrawal Error 0.22 0.22
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3) si v Plant Transient (psig) (psig) Response MSIV Closure 1316 133C Figure 7 (Flux Scram)
- 13. STABILITY ANALYSIS RESULTS (5.4)
Rod Line Analyzed: Extrapolated Rod Block Figure 8 Decay Patio: 0.59 1 Reactor Core Stability Decay Ratio, x2 /*0*
Channel Hydrodynamic Performance Decay Ratio, x2 /*0 -
8x8 Channel: 0.22 P8x8R Channel: 0.18 l
l 4 I
Y1003J01A28 R:v. 0
- 14. ROTATED BUNDLE ERROR RESULTS (5.5.4)
Variable Water Cap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Initial MCPR Resulting MCPR Resulting LHGR (kW/f t) 1.22 1.07 17.67
- 15. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Maximum incremental control rod worth: 0.70% Ak
- 16. LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (5.5.2)
See '?.oss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station," August 1977, NEDO-21696, as amended.
5/6
Y1003J01A28 Riv. 0
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Y1003J01A28 Rev. 0 1 25 l
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Figure 8. Reactor Core Decay Ratio versus Power 14
Y1003J01A28 Rev. O APPENDIX A LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(NEW PROCEDURE)
The Local Rod Withdrawai Error results are reported below in ancordance with Letter, R. E. Engel (GE) to T. A. ippolito (NRC), " Change in General Electric Methods for Analysis of Control Rod Withdrawal Error," May 18, 1981.
ACPR Rod Block Reading
- A8/P8x8R 104 0.13 105 0.16 106 0.19 107* 0.22 108 0.28 109 0.32 110 0.36
- Indicates set point selected.
15/16 (FINAL)
1 G E N E R A L h 'E LE CTRIC