ML19210E283
| ML19210E283 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/30/1979 |
| From: | Henrikson P NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML112920534 | List: |
| References | |
| 79NED405, NEDO-24221, NUDOCS 7912040249 | |
| Download: ML19210E283 (16) | |
Text
NEDO-24221 OCTOBER 1979 f_ k $ k O ) 0kk$
SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR MONTICELLO NUCLEAR GENERATING PLANT RELOAD 7 5
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NEDO-24221 79NED405 Class I October 1979 SUPPLEMENTAL RELOAD LICENSING. SUBMITTAL FOR MONTICELLO NUCLEAR GENERATING PLANT RELOAD 7 P. H. Henrikson Sr. Licensing Engineer Approved: /
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(/ Reload Fuelensing Lic%er 1470 116 NUCLE AR ENERGY PROJECTS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC
NEDO-24221 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT ,
PLEASE READ CAREFULLY This report was prepared by General Electric solely for Northern States Power Company (NSP) for NSP's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending NSP's operating license of the Monticello Nuclear Generating Plant.
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 undertakings of the General Electric Company respecting information in this document are contained in the contract between Northern States Power Company and General Electric Company for nuclear fuel and related services for the nuclear system for Monticello Nuclear Generating Plant, dated December 4, 1967, 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 Company nor any of the contributors to this document makes any representation or war-ranty (express or implied) as to the completeness, accuracy or usefulness of the infomation 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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- 1. PLANT-UNIQUE ITEMS (1.0)*
Plant parameter changes - see Appendix A GETAB initial conditions - see Appendix A Loading Error LHGR -
see Appendix A
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Type Number Number Drilled Irradiated 8DB262 192 None Irradiated 8DB250 28 None Irradiated 8DB219L 52 None Irradiated 8DRB265L 52 52 Irradiated 8DRB282 60 60 New P8DRB282 56 56 New P8DRB265L 44 44 Total 484 212
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core exposure: 15.65 GWd/t. Nominal core average exposure at end of cycle 17.29 GWd/t including coastdown. Core loading pattern: Figure 1.
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND C0hfROL SYSTEM WORTH -
NO VOIDS, 20 C (3.3.2.1.1 AND 3.3.2.1.2)
BOC k gg Uncontrolled 1.114 Fully Controlled 0.955 Strongest Control Rod Out 0.988 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, Ak 0.000
- ( ) refers to areas of discussion in Generic Reload Fuel Application,"
NEDE-240ll-P-A-1, Auguct 1979.
1 1470 118
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) ppm Shutdown (Margin (ak)
(20 C, Xenon Free) 900 0.092
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS '(3.3.2.1.5 AND 5.2)
EOC 8 Void Coefficient N/A* (c/% Rg) -6.98/-8.73 Void Fraction (%) 37.37 Doppler Coefficient N/A (c/ F) -0.228/-0.216 Average Fuel Temperature ( F) 1157 Scram Worth N/A ($) -38.0/-30.4 Scram Reactivity versus Time Figure 2
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)
EOC 8 Exposure 8x8 8x8R P8x8R Peaking factor 1.22, 1.53, 1.40 1.20, 1.66, 1.40 1.20, 1.63, 1.40 (local. radial, axial)
R-Factor 1.098 1.051 1.051 Bundle Power (MWt) 5.157 5.595 5.499 Bundle Flow (10 lb/hr) 105.1 102.7 102.7 Initial MCPR 1.41 1.41 1. 4 '4
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
None
- N = Nuclear Input Data A = Used in Transient Analysis 1470 119 2
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1) 4 Core Power Flow $ Q/A SL Plant Transient Exposure (I) (1, (2 NBR) (2 NBR) (psig) (psig) 8x8/8x8R/P8x8R Responar.
Turbine Trip EOC 8 100 100 411.9 119.7 1177 1215 0.34/0.34/0.37 Figure 3 without Bypass Loss of 100*F -
100 100 117.7 117.0 1022 1067 0.16/0.16/0.17 Figure 4 Feedvater Heater Feedvater EOC 8 100 100 301.4 118.8 1154 1195 0.30/0.30/0.32 Figure j Controller Failure
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(5.2.1)
ACPR LHGR Rod Block Rod Position 8x8/8x8R 8x8/8x8R Limir.ing Reading (Feet Withdrawn) and P8x8R and P8x8R Rod Pattern 104 3.5 0.09/0.20 11.8/13.7 105 3.5 0.09/0.20 11.8/13.7 Figure 6 106 4.0 0.11/0.21 11.9/14.4 107 4.5 0.12/0.23 11.9/14.6 108* 6.0 0.18/0.27 12.4/15.4 109 7.0 0.20/0.29 12.7/15.8
- 11. OPERATING MCPR LIMIT (5.2, Appendix C) 8x8 8x8R P8x8R 1.42 1.46 1.46
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Flow s1 y Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 100 100 1235 1254 Figure 7 (Flux Scram)
- Indicates setpoint selected.
1470 120 3
- 13. STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio: Figure 8 Reactor Core Stability:
Decay hatio, x /*
2 o 0.550 (Natural Circulation-100% Rod Line)
Channel Hydrodynamic Performance Decay Ratio (Natural Circulation-100% Rod Line) 8x8 Channel 0.10 8x8R Channel 0.06 P8x8R Channel 0.06
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
Fuel types 8DRB265L and 8DRB282 were introduced in Reload-6. The Reload-6 MAPLHGR, PCT and Local Oxidation Fraction are applicable to P8DRB265L and P8DRB282, respectively, for Reload-7.
- 15. LOADING ERROR RESULTS (5.5.4)
Limiting event: Mislocated Bundle MCPR 8x8R Monitored 8x8 Mdslocated 1.07
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Maximum incremental control rod worth: 0.97% Ak 1470 121 4
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- 8D8262 E = BDRB282 8 = 808250 F = P80RB282 C = 8DB219L G
- P80RB265L D = 80RB2ML
- Figure 1. Reference Core Loading Pattern
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NEDO-24221 02 06 10 14 18 22 26 30 51 14 47 24 43 10 6 39 22 36 36 35 8 8 8 31 30 36 36 36 27 8 40 0 NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SfMMETRIC, UPPER LEFT QUADRANT SHOWN ON MAP
- 2. NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN R0D
- 3. ERROR ROD AT 22,27 Figure 6. Limiting RWE Rod Pattern 1470 127 10
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NEDO-24221 APPENDIX A PLANT PARAMETER CHANGES Safety / Relief Valve - (Table 5.4, page 5-62, Operating Plants Pressure Relief Systems) 8 S/R valves installed 7 S/R valves assur 1 in analysis capacity at setpoint 82.6% Lowest setpoint = 1108 + 1% psig GETAB Initial Conditions (Table 5-8, page 5-66) Reactor Core Pressure 1038 psia Inlet Enthalpy 524.3 Beu/lb Loading Error Results (5.5.4, Table 5-8, page 5-66) Limiting Event for LHGR: 8x8 Fbnitored 8x8 Mislocated 18.67 kW/ft 1470 130 O 13}}