ML19305D960
| ML19305D960 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/31/1980 |
| From: | Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19305D958 | List: |
| References | |
| 80NED253R, NEDO-24235, NEDO-24235-R1, NUDOCS 8004170020 | |
| Download: ML19305D960 (35) | |
Text
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ . ___ __. ___ _ ____ . ___.
1 RE ON 80NED253R MAR 1980 1
i SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 3 l
GENER AL h ELECTRIC 8004170 43 i 026'
NED0-24235 Revision 1 30NED253R Class I March 1980 l
l SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 3 l
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Prepared:
jj/
P
^
/
[L. Rash Senior Engineer Fuel and Services Licensing Approved: h5( a l R.E.Enge1[ Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENERAL $ ELECTRIC
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IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Carolina Power and Light Company (CP&L) for CP&L's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending CP&L's operating license of the Brunswick Steam Electric Plant Unit 1. 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 Fuel Contract Supplemental A6reement between Carolina Power and Light Company and General Electric Company for Brunswick 1 & 2 dated January 28, 1974, and nothing contained in this document shall be construed as changing said contract. The use of this infcrmation 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 repre-sentation or warranty (express or implied) as to the completeness, accuracy or usefulness of the haformation 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 rssult from such use of such information.
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- 1. ?L.CT-UNIQUE ITEMS (1.0)*
I Transient Analysis Initial Condicions: Appendix 3 i
New Bundle Loading Error Analyses Procedures: Appendix D Linear Heat Generation Race for Sundle 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 Irradiated Inicial Core Type 1 64 64 Initial Core Type 3 88 88 Replacement 7DB230 4 4 Reload 1 SDB274L 100 100 Reload 18DB274H 40 40 Reload 2 SDR3265H 64 64 Reload 2 8DR3283 68 68 New Reload 3 P8DR3265H 132 132 Total 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure: 12,755 mwd /t Assumed reload cycle exposure: 14,943 mwd /t Core loading pattern: Figure 1
- ( ) refers to areas of discussion in NEDE-240ll-P-A, " Generic Reload Fuel Application," August 1979.
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- TE"0-24235 1 CALCULATED CORE EFFECTIW. MULTI?LICATION AND CONTROL 3YSTEM WO RTH - NO V0 EDS , 20'C (3.3.2.1.1 and 3.3.2.1.2) 30C k ..
er:
Uncontrolled 1.097 Fully Controlled 0.939 Strongest Control Rod Out 0.976 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, ak 0.007
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (ak) ppm (20*C, Xenon Free) 600 0.0465
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)
ECC4 E0C4
'E0C4 -1000 mwd /t -2000 mwd /t Void Coefficient -8.30/-10.99 -9.18/-11.47 -9.11/-11.38 N/A* (c/% Rg)
Void Fraction (%) 41.76 41.76 11.76 Doppler Coefficient -0.227/-0.215 -0.221/-0.210 -0.215/-0.204 N/A (c/*F)
Average Fuel 1356 1356 1356 Temperature (*F)
Scram Worth N/A (S) -38.88/-31.10 -37.59/-30.07 -37.25/-29.80 Scram Reactivity Figure 2a Figure 2b Figure 2C versus Time
- N = Nuclear Input Data A = Used in Transient Analysis 2
NED0-24239
- 7. ?ILOAD-UNIQUE GET.G IRANSIENT ANALYS IS INITIAL COND*T!CN ?ARAMETERS (5.2) 7t7 3x3 3x3R ?1x3R 50C'. IOC; 10C; EOCi IOC4 ICCi EOC4 IOC4
-LOOO 2000 .L000 2000 .;000 2000 .LOOO 2000
- 0C'. TJd/: ?idi: IOC; Mid/t %d/ : Eoca .T4a / : %'d / ; COC; %d / c :4d / :
Possing Factor L . ;'. , L.24, 1.24, L.22. L.22. L.22, t.20 . 20 L.20. i.20 1.10. i.20 (local. rastal. 4x1411 L.21. L.23. . 29 1. 30 L. 37 L .14 L. 3 L.32. i.37 i.et. L.50. 1.38 L.10 . a0 L. 0 1.10 L. 0 L. 0 1.;0 1.40 L. 0 L. 0 L. 0 i. 0 i fac:or . 100 L.100 L.100 1.098 1.098 L.098 1.03L L.05L L.051 1.0$L L.051 L.051 Sundle Power (Wc) 3.172 5.433 5.5L3 3.529 5.359 9.133 6.037 6. 35 o. ra7 6.009 6.372 6.722 3undle Flow 127.3 L25.o 125.2 Ll6.3 L14.6 LL2.3 tie.+ L t4. 7 113.. 117.6 L15 . 4 113.a (103 lb/he)
Initial MC?R L.23 !.21 1.11 L.35 L.27 L 20 1.34 L.26 L.21 1.36 L.23 L.20
- 8. SELECTED MARGZN IMPROVEMENT OPTIONS (5.2.2)
Ther=al Power Monitor Exposure Dependent Limi:s
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Power Flow i 4/A Psi Pv Planc Transient Exoosure (2) (2) (2) ( P.) (psig) (psig) 7x7 3x3 3xdR ?8x5R Response Load Rejection ECC4 104 100 342 120 1176 1223 0.21 0.23 0.27 0.29 Figure Ja No 3ypass Loaa Rejection EOC4 104 100 239 115 1170 1217 0.i3 0.19 0.19 0.21 Figure 3b i No Sypass -1000 l Wd/ c t.oad Rejection EOC4 104 100 195 107 1164 1209 0.03 0.07 0.07 0.08 Figure 3c No Sypass -2000 i
t Wd/ t Loss of 100'T 30C4 104 100 124 122 <1100 <1100 0.12 0.14 0.14 0.14 Figure &
Feedvacer EOC4 Heater Feedvace r EOC4 104 LOO 117 110 1044 1091 0.05 0.06 0.06 0.06 Figure $a Controller Failure Feedvacer EOC4 104 .00 119 110 1941 1088 0.05 0.06 0.06 0.07 Figure $b cont rolle r -1000 Failure Wd/ t Feedvater EOC4 104 100 113 110 1040 1087 0.05 0.06 0.06 0.07 Figure 5c Controller -2000 Failure Wd/ t 3
NED0-24235
- 10. LOCAL 10D WI HDRAWAL ERROR (WITM LIMITING INSTRUMENT ?AILURE)
- RANSIENT y
SUMMARY
(5.2.1)
Poftcion EsGR (kW/fe)
Rod Stock (Feet ;CPR 3x3R/ 3xdR/ Limiting Reading '4 L:hd rawn) 7x7 3x3 73xd1 7x7 3x3 ?$x3R Rod ?acteen 104 3.3 0.12 0.17 0.09 16.5 15.1 16.0 Figure 6 105 4.0 0.14 0.21 0.11 17.3 16.L L7.3 Figure 6 106* 4.0 0.14 0.21 0.11 17.3 16.1 17.3 Figure 6 107 4.5 0.16 0.24 0.13 17.5 L6.5 17.F Figure 6 108 5.0 0.18 0.27 0.16 17.7 16.3 17.9 Figure 6 i
LO9 5.0 0.13 0.27 0.16 17.7 16.3 17.9 Figure 6
- 11. OPERATING MCPR LIMIT (5.2) 30C4 to EOC4 EOC4 -2000 Wd/t to EOC4 -1000 Wd/t
-2000 Wd/t E0C4 -1000 Wd/t to EOC4 1.21 1.21 1.28 (7x7 fuel) 1.28 1.28 1.35 (3x8 fuel) 1.21 1.26 1.34 (8x8R fuel) l 1.21 1.28 1.36 (P8x8R fuel)
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! 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Flow si v Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 10 4 100' 1238 1268 Figure 7 (Flux Scram) j 13. STABILITY ANALYSIS RESULTS (5.4) l l Decay Ratio: Figure 8 l
Reactor Core Stability:
Decay Ratio, x2 /x0 0.75 (105% Rod Line -
Natural Circulation l'
' Power) 1 l Channel Hydrodynamic Performance Decay Ratio, x2 /x0 Extrapolated Rod Block Line - Natural Circulation Power) 7x7/8x8/8x8R channel 0.16/0.28/0.21
- Indicates set point selected 4
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS. (5.5.2)
P3DR3255H Exposure MAPLEGR ?CT Location Oxidation (mwd /t) (kW/ft) (*F) Fraction 200 11.5 2138 0.028 1000 11.6 2146 0.028 5000 11.9 2174 0.030 10,000 12.1 2187 0.031 15,000 12.1 2196 0.032 20,000 11.9 2177 0.030 25,000 11.3 2101 0.024 i 30,000 10.7 2016 0.018 I
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- 15. LOADING ERROR RESULIS (5.5.4)
Limiting Event: Rotated PSDR3265H MCPR: 1.07
- 16. CONTROL R0D DROP ANALYSIS RESULTS (3.5.1)
Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Fur.ction: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 i
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RELOAD 1808274L 8 = NOT USED P = RELOAO 1808274H l C =
INITI AL CORE TYPE 3 G= RELOAO 2 80RB26SH O= R EP t.ACEV ENT 708200 H = RELOAD 2 30R8283
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NED0-24235 02 06 10 14 18 22 26 51 10 12 47 43 2 6 8 39 35 10 6 26 0 31 36 27 12 2 12 26 i
Notes: 1. Rod Pattern Is 1/4 Core Mirror Symmetric, Upper Left Quadrant Shown on Map.
i 2. Numbers Indicate Number of Notches Withdrawn out of'48. Blank Is a Withdrawn Rod.
- 3. Error Rod Is 26-35.
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NED0-21235 1.2 ULTIMATE PERFORMANCE LIMIT 1.0 - - - -
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NED0-24235
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Figure 9. Doppler Reactivity Coefficient Comparison for RDA
! 20
NEDO-24233 200 160 -
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- IEDO-26235 I
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NED0-26233 70 s0 -
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NED0-26235 APPENDIX A NOT USED A-1/A-2
NEDO-24235 A?PENDIX 3 TRANSIENT ANALYSIS INITIAL CONDITIONS S/RV Loves: Seepoint (psig) 1105 ' 1*
S/RV Capaci:y (*) 37.a Turbine Pressure (psis) 975.1 1
l B-1/B-2
NEDO-26239 APPE:IDIX C NOT USED c-1/c-2
NEDO-21235 APPENDIX D NEW BUNDLE LOADING ERROR IVENT ANALYSES PROCEDURES The bundle loading error analyses results presented in Section 15 in this supplement are based on new analyses procedures for both the rotated bundle and the mislocated bundle loading error events. The use of these new anal-ysis procedures is discussed below.
NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error analysis results presented in this supple-ment are based on the new analysis procedure described and approved in Ref-erence D-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 utilizes a variable water gap which is more representative of the actual condition, since che interiacing between the top guide and the fuel spacer buccons, 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 calculacion was performed using the same analytical models as were previously used. The only l change is in the simulation of the water gap, which more accurately represents l the actual geometry. l 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 machod of performing the analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core.
D-1
l NEDo-34235 The use of the scaciscically corrected Haling analyses procedure indicaces that the minimum CPR for mislocaced bundles is greater chan che safecy limic (1.07) for all exposures chroughout the cycle.
REFERENCES D-1. Safecy Evaluacion Report (leccer), D.G. Eisenhuc (NRC) to R.E..Engel (GE), ;
MFN-200-78, dated May 8, 1978. ;
A S
- l P
4 1
D-2
NEDO-2635 APPENDIX E LINEAR HEAT GENERATION RATE FOR 3UNDLE LOADING ERROR 16.3 kW/ft E-1/E-2
NED0-24233 APPENDIX F DENSIFICATION POWER S?IKING Reference F-1 documents the NRC scaff position chat ". . . it (is) acceptable to remove che 8x3 and 3x8R spiking penalty factor from the plane Technical Specification for those operating 3WR's for which it can be shown that the predicted worst case maximum transienc LEGR's, when augmenced by the power spike penalty, do noc violace che exposure-dependent safecy limic LHGR's."
The BSEP-2 Reload-3 submiccal contains the required information to demon-strace that the stated criterion is mac for BSEP-2, Reload 3. Seccion 10 Rod Withdrawal Error, and Appendix E (Linear Heat Generation Race for Bundle Loading Error) include che densification effect in the calculaced LHGR of i che 8x3 fuels.
REFERENCE i
F-1 "Safecy Evaluacion of the General Electric Machods for the Consideracion of Power Spiking Due to Densification Effects in 3WR 8x8 Fuel Design and Performance," Reaccor Safecy Branch, DOR, May 1978.
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