ML20214Q147
| ML20214Q147 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/23/1986 |
| From: | Charnley J, Plotycia G, Shirley N GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20214Q102 | List: |
| References | |
| 23A4800, 23A4800-R, 23A4800-R00, NUDOCS 8706040255 | |
| Download: ML20214Q147 (18) | |
Text
-.
23A4800 Revision 0 Class I December 1986 l
GENERAL ELECTRIC BOILING WATER REACTOR SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PILGRIH NUCLEAR POWER STATION g
RELOAD 7
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4 N. C.
h riey P. A bert Verified:
A*
- 1 2 4
G.
D'. Piotycia /
Approved:
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- 2 3
G J. S. CTiarnley/f Manag@
Fuel Licensing NUCLEAR ENERGY BUSINESS OPERATIONS
- GENERAL ELECTRIC COMPANY SAN JOSE, CAllFORNIA 95125 1
GENERAL h ELECTRIC 1/2
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23A4800 Rev. 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 provided to General Electric at the time this report was prepared.
The only uadertakings of the General Electric Company respecting informa-tion in this d cument are contained in the contract between Boston Edison Company and General 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 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.
l 3/4 L
23A4800 Rev. O ACKNOWLEDGMENT The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Submittal, were performed by E. G.
Thacker II, of the Nuclear Fuel Engineering Department.
5/6
23A4800 Rev. 0 1.
PLANT UNIQUE ITEMS (1.0)*
GENEISIS Physics Used with GEMINI ODYN Appendix A: Increased Core Flow Throughout the Cycle Appendix B: Initial Conditions:
2.
RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Designation Cycle Loaded Number Irradiated P8DRB282 5
24 P8DRB26511 6
60 P8DRB282 6
112 BP8DRB282 7
32 P8DRB282 7
160 New BP8DRB300 8
192 Total 580 3.
REFERENCE CORE LOADING PATTERN (3.3.1) mwd /MT mwd /ST Nominal previous cycle core average exposure at end of cycle:
18362 16658 Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:
18362 16658 j
Assumed reload cycle core average exposure at end of cycle:
20593 18682 Core loading pattern:
Figure 1
- ( ) Refer to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, May 1986. A letter "S" preceding the number refers to the U.S. Supplement, NEDE-24011-P-A-8-US, May 1986.
7
23A4800 Rev. 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)
Minimum Shutdown Margin, BOC, k,ff Uncontrolled 1.116 Fully Controlled 0.962 Strongest Control Rod Out 0.990 R, Maximum Increase in Cold Core Reactivity with 0.000 Exposure into Cycle,ak 5.
STANDBY LIQUID CONTROL SYSTEM SHUIDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (ak) ppm (20*C, Xenon Free) i 700 0.044 l
6.
RELOAD UNIQUE TRANSIENT ANALYSIS INPUI (3.3.2.1.5 AND S.2.2)
(COLD WATER INJECTION EVENTS ONLY)
Void Fraction (%)
36.2 Average Fuel Temperature (*F) 1147 Void Coefficient N/A* (4/% Rg)
-5.76/-7.20 Doppler Coefficient N/A* (d/*F)
-0.175/-0.166 Scram Worth N/A* ($)
- N = Nuclear Input Data, A = Used in Transient Analysis
- Generic exposure independent values are used as given in " General Electric Application for Reactor Fuel," NEDE-24011-P-A-8-US, May 1986.
8
l 23A4800 Rev. 0 7.
RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETER (S.2.2)
Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt)
(1000 lb/hr)
MCPR BOC8 to BOC8+8282 mwd /MT (B0C8+7513 mwd /ST)
BP/P8x8R 1.20 1.78 1.40 1.051 5.982 105.2 1.30 BOC8+8282 mwd /MT (B0C8+7513 mwd /ST) to EOC8 BP/P8x8R 1.20 1.69 1.40 1.051 5.685 107.1 1.38 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization:
No Recirculation Pump Trip:
No Rod Withdrawal Limiter:
No Thermal Power Monitor:
No Measured Scram Time:
No Number of Exposure Points:
2 9.
OPERATING FLEXIBILITY OPTIONS (S.2.2.3)
Single Loop Operation:
Yes Load Line Limit:
Yes Extended Load Line Limit:
Yes Increased Core Flow:
Yes Flow Point Analyzed:
107.5%
Feedwater Temperature Reduction:
No ARTS Program:
No Maximum Extended Operating Domsin:
No 9
23A4800 Rev. 0
- 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
Flux Q/A ACPR Transient
(% NBn)
(% NBR)
BP/P8x8R Figure Exposure: BOC8 to EOC8 Loss of Feedwater Heating 117 114 0.14 2
Exposure: BOC8 to BOC8+8282 mwd /MT (BOC8+7513 mwd /ST)
Load Rejection w/o Bypass 471 118 0.23 3
Feedwater Controller Failure 257 115 0.16 4
Exposure: BOC8+8282 mwd /MT (BOC8+7513 mwd /ST) to EOC8 Load Rejection w/o Bypass 561 123 0.31 5
Feedwater Controller Failure 317 122 0.25 6
11.
LOCAL R0D WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(S.2.2.1)
(Generic Bounding Analysis Results)
ACPR Rod Block Reading (%)
BP/P8x8R 104 0.13 105 0.16 106 0.19 107 0.22 108 0.28 109 0.32 110 0.36 Setpoint Selected is: 107 10
23A4800 Rev. 0 12.
CYCLE MCPR VALUES (S.2.2 and S.2.5.4)
Non-Pressurization Events:
Exposure Range: BOC8 to EOC8 BP/P8x8R Loss of Feedwater Heating 1.21*
Fuel Loading Error 1.26 Rod Withdrawal Error 1.29 Pressurization Events:
Option A Option B BP/P8x8R BP/P8x8R Exposure Range: BOC8 to B0C8+8282 mwd /MT (B0C+7513 mwd /ST)
Load Rejection Without Bypass 1.36 Feedwater Controller Failure 1.28 Exposure Range: B0C8+8282 mwd /MT (B0C8+7513 HWd/ST) to EOC8 Load Rejection Without Bypass 1.44 1.39 Feedwater Controller Failure 1.37 1.29 13.
OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3)
P,y Py Transient (psig)
(psig)
Plant Response MSIV Closure 1299 1315 Figure 7 (Flux Scram)
- 0ption B is not available for this exposure.
11
23A4800 Rev. 0 14.
LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Event ACPR Misoriented 0.19 15.
CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
Banked Position Withdrawal Sequence has been implemented at the Pilgrim Nuclear Power Station, so the Control Rod Drop Accident Analysis is not required. NRC approval is documented in NEDE-24011-P-A-8-US, May 1986.
16.
STABILITY ANALYSIS RESULTS (S.2.4)
GE BWR 2/3 plants are exempt from performing cycle specific stability analyses. NRC approval is documented in NEDE-24011-P-A-8-US, May 1986.
17.
LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (S.2.5.2)
See " Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station," August 1977, NEDO-21696, as amended.
12
23A4800 Rev. 0 M BiBEBEM M BE BEBEM BEM M M M M BEBEBEBsBEBsBEMBEMBE
- MBEBEBEBEBEBEBEBEMBEMM
- BiBEBEBEMBEBBBEMMBIBBBE
:BsBEBEBEBEBEBEBBBEBEBEMBE
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'::MMBEMMMBEMMBEBEBEM M MBEi+EBEBEBEBEMBEM MEMMMMMMM BEMBBBBBEMM IIIIIIIIIIIIII 1 3 5 7 9111315171921232527293133353739414345474951 FUEL TYPE B = P8DRB265Fi E = BP8DRB282 C = P8DRB282 P = BP8DRB300 Figure 1.
Reference Core Loading Pattern 13
23A4800 Rev. O i
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Plant Response to Loss of Feedwater Heating (B0C8 to EOC8) 14
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15 I
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16
23A4800 Rev. 0 3 VESSEL PRESS Al5E(PSI)
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Plant Renponse to Generator Load Rejection Without Bypass (B0C8&8282 mwd /HT (B0C8+7513 HWd/ST] to E0C8) 17
23A4800 Rev. O is...
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Plant Response to Feedwater Controller Failure (BOC8+8282 mwd /MT [BOC8&7513 mwd /ST) to EOC8) 18
23A4800 Rev. O I
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Plant Response to MSIV Closure (Flux Scram) 19/20 L.-
23A4800 Rev. O APPENDIX A INCREASE CORE FLOW THROUGHOUT CYCLE The analyses performed for Cycle 8 included increased core flow throughout the cycle. There are no concerns regarding reactor internals pressure drop or flow-induced vibration as discussed in the increased core flow analysis document for the EOC-6 (NEDO-30242).
The flow-biased instrumentation for the rod block monitor should be signal clipped for a setpoint of 107%, since flow rates higher than rated would otherwise result in a LCPR higher than reported for the rod withdrawal error.
21/22
23A4800 Rev. O APPENDIX B INITIAL CONDITIONS Transient Analysis Conditions:
Initial Condition Parameter Analysis Value NEDE-24011 value Core Flow, M1b/hr +0.2%
74 69 GETAB Analysis Initial Conditions:
Initial Condition Parameter Analysis Value NEDE-24011 value Core Flow, M1b/hr 74 69 Reactor Core Pressure, psia 1067 1057 Inlet Enthalpy, Btu /lb 528.4 526.1 l
l 23/24 (FINAL)