ML19260C366
| ML19260C366 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/30/1979 |
| From: | Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19260C362 | List: |
| References | |
| 79NED408, NEDO-24224, NUDOCS 7912260246 | |
| Download: ML19260C366 (25) | |
Text
{{#Wiki_filter:"? 1;': NOVEMBE 1979 SUPPLEMENTAL ~ RELOAD LICENSING SUBMITTAL FOR PILGRIM NUCLEAR POWER STATION UNIT 1 RELOAD 4 b5$ EIR 1617 188 GENER AL h ELECTR 7 9122 601b
NEDO-24224 79NED408 Class I November 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITIAL FOR PILGRIM NUCLEAR POWER STATION UNIT 1 RELOAD 4 Prepared:
.L. sh enior Engineer Fuel and Services Licensing .
Approved: R. E. Enge , Manager tb17 t89 Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE CALIFORNIA 95125 GENER AL $ ELECTRIC
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 The amending BECo's operating license of the Pilgrim Nuclear Power Station. 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 Boston Edison Company and General Electric Company for nuclear fuel and related services for the nuclear system for Pilgrim Nuclear Power Station, dated July 114, 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 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 infonnation 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. 1617 190
NEDO-24224
- 1. PLANT-UNIQUE ITEMS (1.0)e Margin to Opening of Unpiped Spring Safety Valves: Appendix A GETAB Analysis Initial Conditions: Appendix B ATWS Recirculation Pump Trip: Appendix C New Bundle Loading Error Analyses Procedures: Appendix D Linear Heat Generation Rate for Bundle Loading Error: Appendix E Densification Power Spiking: Appendix F
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 and 4.0)*
Fuel Type Number Number Drilled 60 0 Irradiated BDB262 124 124 8DB219H 8DB219L 212 212 P8DRB265L 120 120 New P8DRB282 64 64 580 520 Total 3 REFERENCE CORE LOADING PATTERN (3.3.1) Nominal previous cycle exposure: 11,700 mwd /t Assumed reload cycle exposure: 13,910 mwd /t Core loading pattern: Figure 1
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS. 200C (3.3.2.1.1 and 3 3 2.1.2)
BOC k,pp Uncontrolled 1.089 Fully Controlled 0.929 Strongest Control Rod Out 0.967 lfj[ jgl R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, Ok O.011
*( ) refers to areas of discussion in Reference 1.
1
NEDO-24224
- 5. STANDBY LIQUID CONTROL SYST9H SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) ppm (200C, Xenon Free) 700 0.0682
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)
EOCS Void Coefficient N/A' ((/% Rg) -6.08/-7.60 Void Fraction (%) 37.1 Doppler Coefficient N/A ((/5 0F) -0.226/-0.217 Average Fuel Temperature (OF) 1197 Scram Worth N/A ($) -38.53/-30.82 Scram Reactivity Figure 2
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 8x8 P8x8R EOC5 EOC5 Peaking Factors (local, radial and axial) 1.22/1.73/1.40 1.20/1.87/1.40 R-Factor 1.098 1.052 Bundle Power (MWt) 5.832 6.298 Bundle Flow (103 lb/hr) 97.0 97.37 Initial MCPR 1.21 1.22
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
New Bundle Loading Error Analyses Procedures
*N = Nuclear Input Data A = Used in Transient Analysis
}b 7 2
NEDO-24224 9 CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1) Py Plant Power Flow & Q/A P31 6CPR Transient Exposure (%) (5) (5) (5) (psig) (peig) 8x8 P8x8R Response Ganerator Load Rejection w/o Bypass BOC-EOC 100 100 218 107 1259 1272 0.15 0.16 Figure 3 Loss of Feedwater Heating BOC-EOC 100 100 118 117 <1100 <1100 0.15 0.15 Figure 4 Feedwater Controller Failure BOC-EOC 100 100 138 109 1138 1171 0.12 0.12 Figure 5
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
TRANSIENT
SUMMARY
(5.2.1) Rod Position Rod Block (Feet ACPR MLHOR (kW/ft) Limiting Reading Withdrawn) 8x8 P8x8R 8x8 P8x8R Rod Pattern 104 4.0 0.12 0.12 11.6 13.7 Figure 6 105 4.5 0.14 0.14 11.9 13.7 Figure 6 106 5.0 0.15 0.17 22.6 14.1 Figure 6 1078 5.5 0.17 0.19 13.4 14.8 Figure 6 108 6.0 0.19 0.21 13.8 15.2 Figure 6 109 6.5 0.20 0.23 13.8 15.4 Figure 6
- 11. OPERATING MCPR LIMIT (5.2)
BOCS - EOC5 1.29 (8x8 fuel) 1.29 (P8x8R fuel)
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3) Pv Power Core Flow P31 Plant Transient (5) (5) (psig) (psig) Response MSIV Closure (Flux Scram) 100 100 1328 1341 Figure 7 ' Indicates set po nt selected bl[ }h) 3
NEDO-24224 13 STABILITY ANALYSIS RESULTS (5.4) Decay Ratio: Figure 8 Reactor Core Stability: Decay Ratio, x2/XO 0.61 (Extrapolated Rod Block Line - Natural Circulation Power) Channel Hydrodynamic Performance Decay Ratio, x2/X O (Extrapolated Rod Block Line - Natural Circulation Power) 8x8/8x8R channel 0.22
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS,8 (5.5.2) 8DB262 Exposure MAPLHGR PCT Location Oxidation (mwd /t) (kW/ft) (OF) Fraction 200 11.1 2032 0.016 1,000 11.3 2028 0.015 5,000 11.9 2071 0.017 10,000 12.1 2061 0.016 15,000 12.2 2091 0.018 20,000 12.1 2104 0.019 25,000 11.6 2049 0.016 30,000 10.7 1928 0.010 8DB219H Exposure MAPLHGR PCT Location Oxidation (mwd /t) (kW/ft) (OF) Fraction 200 11.2 2038 0.018 1,000 11.3 2032 0.017 5,000 11.8 2056 0.017 10,000 12.2 2102 0.019 15,000 12.3 2131 0.021 20,000 12.1 2128 0.021 25,000 11.3 2015 C 015 30,000 10.2 1866 0.008 1617 194
- 4
NEDO-24224
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS, (5.5.2) (Continued) 8DB219L Exposure MAPLHGR PCT Location Oxidation (Mkd/t) (kW/ft) (OF) Fraction 200 11.4 2039 0.018 1,000 11.5 2039 0.017 5,000 11.9 2064 0.017 10,000 12.1 2098 0.019 15,000 12.3 2126 0.021 20,000 12.1 2126 0.021 25,000 11.3 2013 0.014 30,000 10.2 1866 0.008 P8DRB265L Exposure MAPLHGR PCT Location Oxidation (mwd /t) (kW/ft) (OF) Fraction 200 11.6 2125 0.023 1,000 11.6 2127 0.023 5,000 12.1 2136 0.022 10,000 12.1 2102 0.020 15,000 12.1 2108 0.020 20,000 11.9 2091 0.019 25,000 11.3 2012 0.015 30,000 10.7 1919 0.010 P8DRB282 Exposure MAPLHGR PCT Location Oxidation (mwd /t) (kW/ft) (OF) Fraction 200 11.2 2087 0.020 1,000 11.2 2083 0.020 5,000 11.8 2110 0.021 10,000 12.0 2097 0.020 15,000 12.1 2108 0.020 20,000 11.8 2088 0.019 25,000 11.3 2011 0.015 30,000 11.1 1961 0.012 jf3e
- A MAPLHGR multiplier of 0.95 is required for operation at flow less than 90%
of rated. 1617 195 5
NEDO-24224
- 15. LOADING ERROR RESULTS# (5.5.4)
Limiting Event: Rotated P8DRB282 MCPR: 1.07
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Maximum Incremental Control Rod Worth: 0.95% 4 1617 196 6
NEDO-24224
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- NEDO-24224 02 06 10 14 18 22 26 51 4 26 47 26 32 43 2 39 2 36 0 35 2 6 31 2 10 36 27 2 2 2 2 NOTES: 1. Rod pattern is 1/4 core mirror symmetric upper left quadrant shown on map.
- 2. No. indicate number of notches withdrawn out of 48. Blank is a withdrawn rod.
3 Error rod is 22-39. 1617 202 Figure 6. Limiting RWE Rod Pattern 12 '
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NEDO-24224 1.2 ULTIMATE PERFORMANCE LIMIT i,0 NATURAL CIRCULATION 0.8 - "O 100% ROD LINE 9 k O.6 - e a 8 0 0.4 0.2 - l I I l o 60 80 100 0 20 40 POWER (%) Figure 8. Decay Ratio i 14
NEDO-24224 REFERENCES
- 1. " General Electric Boiling Water Generic Fuel Application," NEDE-24011-P, Revision 3, March 1978.
1617 205
'J t 9
15/16
NEDO-24224 APPENDIX A MARGIN-TO-SPRING SAFETY VALVES The rationale for changing the basis for providing pressure margin to the spring safety valves is presented in: J. F. Quirk (GE) letter to Olan D. Parr (NRC), " General Electric Licens-ing Topical Report NEDE-24011-P-A, ' Generic Reload Fuel Appication,' Appendix D, Second Submittal," dated February 28, 1979. On this basis the plant can operate at full power throughout the cycle. The core response to the limiting infrequent event is given in Table A-1 and Figure A-1. Table A-1 CORE-WIDE TRANSIENT ANALYSIS RESULTS P 31 Pv Power Flow Plant Transient Exposure (%) (%) (psig) (psig) Response MSIV Closure Trip Scram BOC4EOC 100 100 1158 1188 Figure A-1 16171206 A-1
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NEDO-24224 APPENDIX B GETAB INITIAL CONDITIONS Table 5-8 of Reference 1 states the "Nonvarying Plant GETAB Analysis Initial Conditions." The PNPS parameters, core pressure inlet enthalpy, and nonfuel power fraction are given as 1045 psia, 526.1 Etu/lb, and 0.035, respectively. Values of 1065 psia, 526.6 Btu /lb, and 0.030 which more nearly reflect actual plant data, were assumed for this submittal. Reference 1 will be revised to eliminate these discrepancies. 1617 208 B-1/B-2
NEDO-24224 APPENDIX C ATT) RECIRCULATION PUFP TRIP Reference 1 states that PNPS has no ATWS-RPT. BECo will install ATWS-RPT during the fourth refueling outage. The transient analyses described in this document assume the ATWS-RPT is installed and functioning. Reference 1 will be revised to reflect this plant modification. 1617 209 C-1/C-2
NEDO-24224 APPENI ; D NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES The bundle loading error analyses results presented in Section 15 in this supple-ment are based on new analyses procedures for both the rotated bundle and the mislocated bundle loading error events. The use of these new analysis proca-dures is discussed below. NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error analysis results presented in this supplement are based on the new analysis procedure described and approved in Reference D-1. This new metrod of performing the analysis is based on a more accurate detailed analytical model. The principle difference between the previous analysis procedure and the new analysis procedure is the modeling of the water gap along the axial length of the bundle. The previous analysis used a uniform water gap, whereas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interfacing between the top guide and the fuel spacer buttons, caused by misorientation, causes the bundle to lean. The effect of the variable water gap is to reduce the power peaking and the R-factor in the upper regions of the limiting fuel rod. This results in the calculation of a reduced CPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry. NEW ANALYSIS PROCEDURE FOR THE MISLOCATED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analyses results presented in this supplement are based on the new analysis procedure described in Reference D-1. This new method of performing the analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core. The use of the statistically corrected Haling analyses procedure indicates that the minimum CPR for mislocated bundles is greater than the safety limit (1.07) for all exposures throughout Cycle 7. REFERENCES D-1. Safety Evaluation Report (letter), D.G. Eisenhut (NRC) to R. E. Engel (GE), MFN-200-78, dated May 8, 1978. 1617 210 D-1/D-2
NEDO-24224 APPENDIX E LINEAR HEAT GENERATION RATE FOR BUNDLE LOADING ERROR 17.7 kW/ft 1617 2ll E-1/E-2
NEDO-24224 APPENDIX F DENSIFICATION POWER SPIKING Refe"ence F-1 documents the NRC staff position that ". . . it (is) acceptable to remove the 8x8 and 8x8R spiking penalty factor from the plant Technical Specification for those operating BWR's for which it can be shown that the pre-dicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGR's". The PNPS-1 Reload-4 submittal contains the required information to remove the power spiking penalty from the PNPS-1 Technical Specifications. Section 10, Rod Withdrawal Error, and Appendix E (Linear Heat Generation Rate for Bundle Loading Error) include the densification effect in the calculated LHGR of the 8x8 fuels. REFERENCES F-1 " Safety Evaluation of the General Electric Methods for the Consideration of Power Spiking Due to Densification Effects in BWR 8x8 Fuel Design and Performance," Reactor Safety Branch, DOR, May 1978. 1617 212 F-1/F-2
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