ML19209B765
| ML19209B765 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/31/1979 |
| From: | Engel R, Ervin A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19209B762 | List: |
| References | |
| 79NED294, NEDO-24209, NUDOCS 7910100409 | |
| Download: ML19209B765 (26) | |
Text
'
uSEl AUGUS 9 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY NUCLEAR PLANT UNIT 1 RELOAD NO. 3 E00R ORIGK 1139 013 79101 00 O
?
GEN ER AL h ELECTR
NEDO-24209 79NED294 Class I August 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BR0hNC FERRY NUCLEAR PLANT UNIT 1 RELOAD NO. 3 Prepared by / ' /w A.rf.'Ervin' Approved by: C _
R. E. Enge1, # Actir3 Manager Reload Fuel Licensing 1139 ?i4 NUCLE AR ENERGY P3OJECTS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFOHNIA 95125 GEN ER AL @ ELECTRIC
IMPORTANT NOTICE REGARDING CONTENTS OF '9IS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for The Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending TVA's operating license of the Browns Ferry Nuclear Unit 3 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 The Tennessee Valley Authority and General Electric Company for nuclear fuel and related ser vices for the nuclear system for Browns Ferry Nuclear Plant Uni 3, dated June 17, 1966, 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 evah 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 centained 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.
1139 dis
NED0-24209
- 1. PLANT-UNIQUE ITEFS (1.0)*
Fuel Loading Error LHGR: Appendix A Parameters Different from Reference 1: Appendix A Item 3. Revised Reference Core Loading Pattern Format: Appendix B New Bundle Loading Error Event Analysis Procedures: Appendix C Items 7, 9, 10, 11, 13: Format change to include additional 8x8R/P8x8R column Item 13. Stability Analysis Results: Reference 1 (Appendix D)
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 and 4.0)
Fuel Type Number Number Dril.1ed Irradiated Initial Core Type 2 100 100 Irradiated Initial Core Type 3 108 108 Irradiated 8DB274H 24 24 Irradiated 8DB274L 144 144 Irradiated 8DRB265L 88 88 Irradiated 8DRB265H 68 68 New** P8DRB284 2?? 232 TOTAL 764 764
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
See Appendix B.
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYS1EM WORTH - NO VOIDS, 20*C (3.3.2.1, i.e., 3.3.2.1.2)
BOC K gg Uncontrolled 1.122 Fully Controlled 0.961 Strongest Control Rod Out 0.989 R, Maximum Increase in Cold Core Reactivity with 0.000 Exposure Into Cycle, Ak
- ( ) refers to areas of discussion in Reference 1.**
- " General Elect 11c Boiling Water Reactor Generic Reload Fuel Application,"
August 1979, (N EDE-240ll-P-A-1) .
1 1139 016
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (tk) ppm (20*C, Xenon Free) 600 0.027
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)
Void Coefficient N/A,* (c/% Rg) -7.31/-9.135 Void Fraction (%) 40.18 Doppler Coefficient N/A (c/*F) -0.2316/-0.22 Average Fuel Temperature (*F) 1383 Scram Worth N/A ($) -37.1/-29.6 Scram Reactivity Figure 2
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 7x7 8x8 8x8R/P8x8R Peaking factors (local, 1.24 1.22 1.20 radial and axial) 1.31 1.42 1.56 1.40 1.40 1.40 R-Factor 1.10 1.098 1.052 Bundle Power (MWt) 5.517 5.980 6.569 Bundle Flow (10 lb/hr) 120.68 108.07 108.75 Initial MCPR 1.20 1.24 1.24
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
Recirculation Pump Trip (REDY Code Only) 1139 ni7
- N = Nuclear Input Data A = Used in Transient Analysis 2
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Power Core Flow h Q/A SL v aCPR Plant Transient Exposure (%) (%) (% NBR) (% NBR) (PSIC) (PSlus 7x7 8x8 8x8R/P8x8R Respons(
Cenerator Load Rejection EOC4 104.5 100 242 111.1 1199 1227 0.12 0.17 0.18 Figure 3 without Bypass Loss of 100*F j Feedwater 104.5 100 123.7 123.3 1013 1069 0.13 0.15 0.15 Figure 4 Hea'Inc, f Feedwater I Controller EOC4 104.5 100 170.6 112.2 1157 1190 0.08 0.12 0.12 Figure 5 Failure
- 10. LOCAL R0D WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(5.2.1)
Rod Position ACPR ~
LHGR Limiting Rod Block (Feet Rod Reading Withdrawn) 7x7 8x8 8x8R/P8x8R 7x7 8x8 8x8R/P8x8R Pattern 104 3.0 0.10 0.11 0.10 16.0 12.90 14.07 Figure 6 105 3.5 0.15 0.15 0.13 17.56 14.05 14.95 Figure 6 106 3.5 0.15 0.15 0.13 17.56 14.05 14.95 Figure 6 307 4.0 0.24 0.18 0.15 18.57 14.98 16.27 Figure 6 108 4.5 0.28 0.20 0.17 18.75 15.25 16.76 Figure 6 109 4.5 0.28 0.20 0.17 18.75 15.25 16.76 Figure 6 110 5.0 0.31 0.24 0.18 18.84 15.47 17.03 Figure 6
- 11. OPERATING MCPR LIMIT (5.:.)
1.23 (7x7 fuel) 1.24 (8x8 fuel) 1.25 (8x8R/P8x8R)
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Flow sl v Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 104.5 100 1236 1272 Figure 7 (Flux Scram)
- Indicates serpoint selected.
1139 n18 3
- 13. STABILITY RESULTS (5.4)
Decay Ratio: Figure 8 Reactor Core Stability Decay Ratio, x2/*0
- Channel Hydrodynamic Performance Decay Ratio, x2 /*0 7x7 channel 0.262 8x8 channel 0.406 8x8R/P8x8'.< channel 0.327
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
P8DRB284 Exposure MAPLHGR PCT Local Oxidation (mwd /t) (kw/ft) (*F) Fraction 200 11.2 1685 0.004 1000 11.3 1667 0.003 5000 11.8 1671 0.003 10000 12.0 1647 0.003 15000 12.0 1669 0.003 20000 11.8 1672 0.003 25000 11.2 1633 0.003 30000 10.8 1596 0.002
- 15. LOADING FRROR RESULTS (5.5.4)
See Appendix C.
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functic as: Figures 12 and 13 Plant Specific Analysis Results Parameter (s) not bounded: Scram Reactivity Shape Function at 20*C Resultant peak enthalpies (cal /g): 152.63 1139 0i9 4
- s m 58 @dssssssss_sm.sse@ @'@ d@ d-@ @
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'HMMMMMM" III!IIIIIIIIII 01 03 05 07 09 11 13 15 17 10 21 23 75 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 FUEL TYPE A = INITIAL CORE TYPE 2 E = 8DRB265L B = INITIAL CORE TYPE 3 F = 80RB265H C = 8DB274L G = P8DRB284 p g
[DilDitbin'M Figure 1. Reference Core Loading Pattern 5 h2;)
NEDO-24209 100 45 CONTROL ROD DRIVE VS TIME SCRAM REACTIVITY VS TIME 90 -
m_ 4o 67B CRD IN PERCENT A -
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NOMINAL SCRAM CURVE IN (-$ )
SCRAM CURVE USED IN ANALYSIS 70 30 60 -
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Figure 2. Scram Reactivity and Control Rod Drive Specification 1139 021
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ru TIME (SEC) TIME (SECl LN Figure 4. Plant Response to Loss of 100*F Feedwater Heating
1 NEUI'10N FLUX 1VESSELPdESRISE(PSil .
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Figure 5. Plant Response to Feedwater Controller Failure
3.- : ...
NEDU-24209 , ,
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02 06 10 14 18 22 26 30 ,
59 00 00 55 32 51 00 10 00 s
47 34 43 00 18 00 16 39 36 28 26 35 00 02 00 02 31 30 38 38 NOTES:
a
- 1. Rod Pattern Is 1/4 Core Mirro- Symmetric.
- 2. Numbers Indicate Number of Notches Withdrawn.
- 3. Error Rod Is (22-43).
Figure 6. Limiting RWE Rod Pattern 1139 '25 10 A .
1 l
1 NEUIRON F LUX 1 VESSEL PfES HISE (PSin 2 AVE SURffCE HEAT FLUX 2 RELIEF VfLVE FL ChP 3 g cw* 3 CCHE INL[ T f t L1H 300*
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2 VESSEL SlERMFI DW 3 TURBINE c TEAMFLOW 2 DOPPLEN 3 SC VITT FCTIVITY 200* 3*
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- O. 4. 8. 12. 16. O. 0.6 1.2 1.8 2.4 rg TIME (SEC1 TIME ISEC1 GN Figure 7. Plant Response to MSIV Closure
NED0-24209 1.2 L'LTIMATE STABILITY LIMIT 1.0 -- - ------- - - - - -
0.8 -
o N ATUR AL CIRCULATION X
x" 0.6 -
9 e
e k 105% ROD LINE S
O 0.4 -
0.2 I I I I 0.0 O 20 40 60 80 100 PERCENT POWER Figure 8. Decay Ratio 1139 n27 12
NEDO-24209 0 '-
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O BOUNDING VALUE FOR 280 CAL /G. COLD O BCUNDING VALUE FOR 280 CAL /G. HSB A CALCULATED VALUE-COLO Q CALCULATED VALUE-HSB
-35 O 4% 800 1200 1600 2000 2400 FUEL TEMP, DEG C Figure 9. Doppler Reactivity Coefficient Comparison for RDA l139 123 13
NED0-24209 20 16 -
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O BOUNDING VALUE FOR 280 CAL /G.
O CALCULATED VAli)E O
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O 4 8 12 16 ROD POSITION, ft OUT Figure 10. RDA Reactivity Shape Function at 20*C 14
NED0-24209 20 0 BOUNDING VALUE FOR 280 CAL /G O CALCULATED VALUE 16 -
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NEDO-24209 100, O BOUNDING VALUE FOR 280 CAL /G O CALCULATED VALUE 80 -
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0 2 4 6 8 ELAPSED TIME, sec Figure 13. RDA Scram Reactivity at 286*C 17/18 !)30 l32
NEDO-24209 APPENDIX A Fuel Loading Error LHGR* (kW/ft): 18.0 Parameters Different From Reference 1 Number of Safety Valves: 0 Safety / Relief Valve Capacity at Setpoint** (No./%): 12/76.246 GETAB Analysis Initial Conditions Reactor Pressure (psia) : 1035 Inlet Er.thalpy (Btu /lb): 521.5
- Includes 0.02 penalty for R-Factor uncertainty
- 13S/R valves installed, however 1 valve assumed to be out of servict 1139 n33 A-1/A-2
, NEDO-24209 APPENDIX B
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end f CYCIC: 15,377 Wd/t Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 15,377 mwd /t 4
~-
Assumed reload cycle core average exposure at end of cycle: 16,640 mwd /t Core loading pattern: Figure 1 1139 034 .
h B-1/B-2 Pe
~ '
4
NEDO-24209 APPENDIX C NEW BUNDLE LOADING ERROR EVENT NRALYSES PROCEDURES The bundle loading error analyses results presented below are based on new analyses procedures for both the rotated bundle and the mislocated bundle-loading error events. The use of these new analyses procedures is discussed below.
C.1 NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated burdle loading error event analysis results presented in this supplement are based on the new analysis procedure described and approved in Reference C-1. This new method of performing the analysis is based on a more accurate detailed analytical model.
The principal 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 vtilizes a variable water gap which is more repre entative of the actual condition, since the interfacing between the top guide and the fuel spacer buttons, caused by misorientation, causes the bundle te lean. The effect of the variable water sap is to reduce the power peaking and the R-factor in the uppar regions of the limiting fuel rod. This results in the calculation of a redtced CPR for the rotated bundle. The calculation was perforced using the same analytical models as were previously used. The only change is in the simu-lation of the water gap, which uore accurately represents the actual geometry.
Analysis of the most limiting rotated bundle starting from an initial CPR of 1.19 (which includes the 2% allowance for uncertainties as required by the NRC) results in a minimum CPR greater than 1.07.
1i39 '35 C-1
NED0-24209 .
C.2 NEW ANALYSIS PROCEDURE FOR THE MISLOCATED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analyses results presented fu this supplement are based on the new analysis procedure described in Reference B-1.
This new method of performing tl.e analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core.
Tite use of the statistically corrected Haling analyses procedure gives the following results:
Limiting Events: MCPR Rotated 8DB274 21.07 Mislocated 8DRB265 2.1.07 REFERENCES C-1 Safety Evaluation Report (letter, D. G. Eisenhut (NFC) to R. E. Engel (GE), MFN-200-78, dated May 8, 1978.
1139 ":36 C-2
e NUCLEAR ENERGY BUSINESS GROUP
NED0-24209 PUBLICATION NO. ERRATA And ADDENDA SHEET T. I. E. NO.
I TITLE NO.
LICENSING SUBMITTAL W R BFh? September 1979 DATE UNIT 1 RELOAD NO. 3 NO TE: Correct allcopies of the applicable ISSUE DATE September 19 publication as specified below.
REFERENCES INSTRUCTIONS ITEM ,j$ECTION,p PAGE gg , (CORRECTIONS AND ADJITIONS) 01 Page 3 Replace with new page 3.
l
"^ '
1139 337
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Fower Core Flow Q/A 'SL 'v aCPR Flant Transient Exper .e (%) (%' ur' (% NBR) (PSIG) (FSIG) 7x7 8x8 8x8R/P2x8R Respense Generator ]
d Reject EOC4 10s.3 100 .2 111,1 ligg 1227 o,12 o,17 0.18 Figure 3 Less of JO*F '
10J 100 123.7 12*.3 1013 1069 0.13 0.15 0.15 Figure I.
Heating Feedwater Controller EOC4 104.5 100 170.6 112.2 1157 1190 0.09 0.12 0.12 F1 pre 5 Failure
- 10. LOCAL R0D WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(5.2.1)
Rod Position ACPR LEGR Limiting Rod Block (Feet Rod Reading Withdrawn) 7x7 8x8 8x8R/P8x8R 7x7 8x8 8x8R/P8x8R Pattern 104 3.0 0.10 0.11 0.10 16.0 12.90 14.07 Figure 6 105 3.5 0.15 0.15 0.13 17.56 14.05 14.95 Figure 6 106* 3.5 0.15 0.15 0.13 17.55 14.05 14.95 Figure 6 }
107 4.0 0.24 0.18 0.15 18.57 14.98 16.27 Figure 6 108 4.5 0.28 0.20 0.17 18.75 15.25 16.76 Figure 6 109 4.5 0.28 0.20 0.17 18.75 15.25 16.76 Figure 6 110 5.0 0.31 0.24 0.18 18.84 15.47 17.03 Figure 6
- 11. OPERATING MCPR LIMIT (5.2) 1.23 (7x7 fuel) 1.24 (8x8 fuel) 1.25 (8x8R/P8x8R)
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Core Flow sl v Plant Power Transient (%) (%) (psig) (psig) Response MSIV Closure 104.5 100 1236 1272 Figure 7 (Flux Scram)
- Indicates setpoint selected.
1139 338
- - ... - -