ML19259B441
| ML19259B441 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1978 |
| From: | Engel R, Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19259B437 | List: |
| References | |
| 78NED402, NEDO-24170, TAC-07307, TAC-7307, NUDOCS 7902090358 | |
| Download: ML19259B441 (24) | |
Text
{{#Wiki_filter:""fa': DECEMGER 1978 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR COOPER NUCLEAR STATION UNIT 1 RELOAD 4 79020963 C 6 GENER AL h ELECTRIC
NED0-24170 78NED402 Class I Decer::ber 1978 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR COOPER NUCLEAR STATION UNIT 1 RELOAD 4 Prepared'
?'
J. L. Rash Vl Approved: Cd ' ' M'"7.Y ? R. E. Engel, Manager Operating Licenses I NUCLEAR ENE RGY PROJECTS DIVISION
- GENER AL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 93125 G EN ER AL h ELECTRIC
NED0-24170 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ :AREFULLY This report was prepared by General Electric solely for Nebraska Public Pouer District (NPPD) for NPPD's use with the U.S. Nuclear Regulatory Comission (USNRC) for amending NPPD's operating license of the Cooper Nuclear Station. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts knoun, obtained or provided to General Tlectric at the time this report uas prepared. The only undertakings of the General Electric Company respecting information in this document are contained in the contract between Nebraska Public Pouer District and General Electric Company for nuclear fuel and related services for the nuclear eystem for Cooper Nuclear Station 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 pur-pose other than that for which it is intended, is not authorized; and with respect to any such unauthori::ed use, neither General Electric Company nor any of the contributors to this document makes any representation or carranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this documcnt or that such use of such information may not infringe privately ouned rights; nor do they assume any responsibility for liability or damage of any kind uhich may result from such use of such information.
NEDO-24170
- 1. PLANT-UNIQUE ITEMS (1.0)*
Bundle Loading Error Analysis - Appendix A GETAB Analysis Initial Conditions - Appendix B Densification Power Spiking - Appendix C
- 2. RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0)
Fuel Type Number Number Drilled Irradiated Initial Core 7DB250 152 152 Irradiated 8DB250 (R1) 72 72 Irradiated 8DB274L (R162) 112 112 Irradiated 8DB274L (R3) 24 24 Irradiated 8DRB283 (R3) 76 76 New 8DRB283 (R4) 112 112 Total 548 548
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure: 15,753 mwd /t Assumed reload cycle exposure: 16,230 mwd /t Core loading pattern: Figure 1
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH -
NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2) BOC k gg Uncontrolled 1.110 Fully Controlled 0.952 Strongest Control Rod Out 0.986 R, Maximum Increase in Code Core Reactivity with Exposure Into Cycle, ak 0.000
*( ) refers to areas of discussion in " Generic Reload Fuel Application,"
NEDE-240ll-P-A, Revision 0, August 1978, 1
NT90-24170
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) 1331 (20'C, Xenc, Free) 600 0.05o
- 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)
EOC5 Void Coefficient N/A* (c/% Rg) -8.15/-10.19 Void Fraction (%) 40.03 Doppler Coefficient N/A (c/%'F) -0.226/-0.217 Average Fuel Temperature (*F) 1352 Scram Worth N/A ($) -38.78/-31.02 Scram Reactivity vs Time Figure 2
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INI11AL CONDITION PARAMETERS (5.2) 7x7 8x8/8x8R Exposure EOC5 EOC5 Peaking factors (local, (1.24, 1,378, 1.4) (1.19, 1.578, 1.4) radial and axial)
R-Factor 1.08 1.054 Bundle Power (MWt) 5.864 6.703 Bundle Flow 119.85 109.86 (103 lb/hr) Initial MCPR 1.20 1.22
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
None
*N = Nuclear Input Data A = Used in Transient Analysis 2
NEDO-24170
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
A Core $ Q/A p p ACPR Power Flow (% (% sl v 8x8/ Plant Transient Exposure _ _( %) (%) NBR) NBR) (psig) (psig) 7x7 8x8R Response Turbine BOC-EOCS 104 100 226 107 1170 1194 0.09 0.14 Figure 3 Trip Without Bypass Load BOC-EOC5 104 100 248 108 1172 1196 0.10 0.15 Figure 4 Rejection Without Bypass Loss of BOC-EOC5 104 100 124 117 1023 1069 0.13 0.14 Figure 5 100'F Feedwater Heating Feedwater BOC-EOC5 104 100 168 110 1123 1165 0.09 0.13 Figure 6 Controller Failure
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(5.2.1) Rod Position Rod Block (Feet ACPR MLHGR (Kw/ft) Limiting Reading Withdrawn) 7x7 8x8/8x8R [x7 8x8/8x8R Rod Pattern 105* 3.5 0.16 0.10 17.3 16.1 Figurc 7 106 4.0 0.23 0.12 17.5 16.4 Figure 7 107 4.5 0.31 0.13 17.5 16.4 Figure 7 108 5.0 0.33 0.14 17.3 16.2 Figure 7 109 5.5 0.33 0.15 17.2 16.2 Figure 7 110 7.0 0.32 0.19 16.7 15.5 Figure 7
- Indicates setpoint selected 3
NED0-24170
- 11. OPERATING MCPR LIMIT (5.2)
BOC to EOC5 1.23 (8x8/8x8R fuel) 1.23 (7x7 fuel)
- 12. OVERPRESSURIZATION ANALYSIS SLWiARY (5.3)
Power Core Flow sl v Plant Transient (%) (%) (psig) (psig) Response su e pf g 104 100 1237 1276 Figure 8
- 13. STABILITY ANALYSIS RESULTS (5.4.)
Decay Ratio: Figure 9 Reactor Core Stability: Decay Ratio, x2 /*0 (105% Rod Line - Natural Circulation Power) 0.79 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 (105% Rod Line - Natural Circulation Power) 8x8/8x8R channel 0.37 7x7 channel 0.23
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
No new bundle types are added; no ECCS Tech Spec changes will be necessitated by this reload. 4
NED0-24170
- 15. LOADING ERROR RESULTS (5.5.4)
Limiting Event: Rotated Bundle (SDRB283) MCPR: 1.07
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient: Figure 10 Accident Reactivity Shape Functions: Figures 11 and 12 Scram Reactivity Functions: Figures 13 and 14 Plant specific analysis results Parameter not bounded: Accident Reactivity Shape Function, Cold Startup Resultant peak enthalpy: 214 cal /gm 5
NED0-24170 s2 DIED El.00 8 8 ODiB ED 0 so 8 00'8 0; 3D B YB 8'ED 8 ED 8 a @ @ @ @ 0D 8 8 @ 8 8 ED 8 00:00 ED ED @ @ a E0 8 8 00 8 0 8 8 8 E8 8 8 E0 8'8 8 00 8 8 ED a 0 8 8 00 8 8 800 8 8 ED 8 8iB ED 8 0 8 8 8 8 0 42 Y Y ED B 59 @@ 8 ED @@ B 8 8'8 8 8 8 8 8 OD ED 8 a 8 0 8 @ 8 @ SiGD 8_ @ 8_ 8 @ 8 B i B B B @ @ 5 5 >= ED E+D ED EDo 8ED 8'ED 00 8 8 8 0 0 8'UD 8 ED ED 0 8 ED ED as -O GD DiGD BiED EiDD SiED 8 9 BiGD 85 ED 8 8 ED E_D 5 5 5 00 0 Y 24 -ED B 818 00's BD'ED 8'ED ED ED ED'ED EDED ED E B ED 8 00 B ED ED B Y YY 22 -EDiB 8_@ 8i00 8 8 85 8,8 Bi8 8 8 bib OD 8 8 8 58 8 ED so -00'ED 8 8 8'UD ED ED ED 8 8'ED 8'8 8 8 8 OD ED ED 8 ED 8 ED ED ED Y 28 -BiED BiED ED 8 8 GD ED 8 8 8 8 8 8 8 8 ED ED:8 8 28 -00'00 8'8 B BYYYY S ED ED ED B ED E ED ED Y8 ED 00 8'8 @@ED EDY88 ED C 0D0+0 8 24 -ED 8 58 @ @ bib UD ED @@ 8_@ $@ $5 SiED SS B E0 8 8 Y 22 -ED B B B B B 8'ED B YB B B B B B B B B E]'8 OD 8 EDYY O B ED 2o -EDiED 8 8 8 ED BiED @i8 ED E 8 8 8 ED 8 8 EDi8 8 ED 8 ED ED ED is -O'8 B YY O B B 8800 8'8 B YY ED T B @@ Y B B 8'8 BY8 5@ 00 C is BD EDiO 8_ ED 8 ED 095 5@ S@ bib B B ED 8 8 00 BiED ED - 8'00 8'ED B YB 8 8 ED 8 8 8 ED'8 ED B ED 8 Y 0 8 0'8 l l; EDiED 88 0009EDiED8 5B ED Op888 E9 55 85 55 BpD , E0 '2 io ; 4
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- RELOAD 1 AND 2 808274 8 INITIAL CORE 7D8250 F
- RELOAD 3 808274 C = INITIAL CORE 708250 G = RELOAD 3 80RB283 D = RELOAD 1808250 H = RELOAD 4 80R B283 Figure 1. Reference Core Loading Pattern 6
NED0-24170 100 4r., 90 40 80 35 70 30 60 25 g i A 50 D G 8 ' E 678 CRD IN PERCENT 20 $ 40 15 30 10 20 NOMINAL SCRAM CURVE IN (-5) 5 10 SCRAM CURVE USED IN ANALYSIS 0 0 0 1 2 3 4 TIME (sec) Figure 2. Scrara Reactivit y and Control Rod Drive Specifications 7
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Figure 4. Plant Response to Generator Load Rejection, without Bypass
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TIME ISEC1 TIIE (SEC) Figure 5. Plant Response to Loss of 100 F Feedwater IIeating, MFC
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NEDO-24170 02 06 10 14 18 22 26 30 51 47 10 10 43 30 18 39 6 10 6 35 30 40 42 31 14 0 18 27 40 42 42 23 NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC UPPER LEFT QUADR ANT SHOWN ON MAP.
- 2. NUMBERS INDICATE NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.
- 3. EhROR ROD iS 118.31).
Figure 7. Limiting RWE Rod Pattern 12
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Figure 8. Plant Response to MSIV Closure, Flux Scram
NEDO-24170 12 ULTIMATE STABILITY LIMIT _ ,, _ 0.8 ^O X _N O NATURAL Q O.6 CIRCULATION e w O 105% ROD LINE 0.4 02 0 0 20 40 60 80 100 PE RCE NT POWE R Figure 9. Decay Ratio 14
NEDO-24170 0 IN 1/ DELTA degrees C (MILLIONTHS)
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w - 15 2 i 8 O o ~20 0 O SOUNDING VALUE FOR 280 cal /g COLD A BOUNDING VALUE FOR 280 cal /g HS8
- 25 O CALCULATED VALUE - COLD h CALCULATED VALUE - HSB -30 V
i 1 0 400 800 1200 1600 2000 2400 FUEL TEMPERATURE (OC) Figure 10. Fuel Doppler Reactivity Coefficient Comparison for RDA 15
NED0-24170 20 16 G h12 - S 5, d O I N b 28 - U 3 m i O BOUNDING VALUE FOR 280 cal /g O CALCULATED VALUE 4 - i l I l g 0 4 8 12 16 20 ROD POSITION (ft OUT) Figure 11. RDA Reactivity Shape Function at 20*C 16
NED0-24170 20 16 G [ BOUNDING VALUE FOR 280 cal /g S12 O O E v t 5 h8 E CALCULATED VALUE __ _ 3 4 0[_ 0 4 8 12 16 20 ROD POSITION (ft OUT) Figure 12. RDA Reactivity Shape Function at 286*C 17
NEDO-24170 70 60 - s0 G ? O Z $ 40 8 5 N 5m P N a 20 - CALCULATE D yALUE BOUNDING VALUE FOR 280 cal /g 10 k OC ^ 0 4 8 ELAPSED TIME (sec) Figure 13. RDA Scram Reactivity Function at 20*C 18
NED0-24170 100 80 5 s != 2 N I i 2 2 G# CALCULATED V ALUE BOUNDING VALUE FOR 280 cal /g 20 0(2 2 2 0 2 4 6 8 10 ELAPSED TIME bec) Figure 14. RDA Scram Reactivity Function at 286*C 19/20
NEDO-24170 APPENDIX A BUNDLE LOADING ACCIDENT The loading error accidents nislocated bundle and rotated bundle) for Cooper Cycle 5 have been analyzed using the " revised methods" (Reference A-1). These analysis methods have been previously applied to Cooper (References A-2 and A-3) and these applications approved by the NRC (References A-4 and A-5) . The loading errors will not influence the opera t: ng CPR limit for 7x7 fuels; however the analysis of a rotated 8DRB281 bundle indicates more severe conse-quences than any abnormal operational transient. The operating CPR limit for 8x8 and 8x8R bundles will therefore be established by the rotated bundle accident. This limit includes a bias of 0.02 for R factor uncertainties as required by the NRC (Reference A-1). The most severe MLHGR predicted for a misloaded bundle is 16.4 kW/ft. This value includes an allowance of 2.2% for power spiking due to fuel densification (see Appendix C) . REFERENCES A-1 Letter, D.C. Eisenhut (NRC) to R.E. Engel (GE), transmitting Safety Eval-uation Report on "new calculational procedures. . .for the fuel bundle loading error analyses", May 8, 1978. A-2 Letter, J. Pilant (NPPD) to George Lear (NRC), April 14, 1978. A-3 Letter, J. Pilant (NPPD) to T.A. Ippolito (NRC), August 16, 1978. A-4 Letter, George Lear (NRC) to J.M. Pilant (NPP D) , May 2, 1978. A-5 Letter, T.A. Ippolito (NRC) to J. Pilant (NPPD) , Augus t 25, 19/6. A-1/A-2
NFDO-24170 APPENDIX B GETAB INITIAL CONDITIONS Table 5-8 of Reference B-1 states the "Nonvarying Plant GETAB Analysis Initial Conditions". The Cooper core pressure is given as 1045 psia. A value of 1035 psia, which more nearly reflects actual plant data, was assumed for this s ub mit tal . Reference B-1 wi?1 be revised to eliminate this discrepancy. REFERENCES B-1 Licensing Topical Report, " General Electric Boiling Water Reactor, Generic Reload Fuel Application", NEDE-24011-P-A, May 1977. B-1/B-2
NED0-24170 APPENDIX C DENSIFICATION POWER SPIKING Reference C-1 documents the NRC staff position that ". . .it (is) acceptable to remove the 8x8 and 8x8R spiking penalty factor from the plant Technical Specifi-cation for those operating BWR's for which it can be shown that the predicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHCR's". The Cooper Reload-4 submittal contains the required information to remove the power spiking penalty from the Cooper Technical Specifications. Section 10, Rod Withdrawal Erro , and Appendix A (Bundle Loading Accident) include the den-sification effect in the calculated LHGR of the 8x8 fuels. REFERENCES C-1 " Safety Evaluation of the General Electric Methods for the Consideration of Power Spiking Due to Densification Ef fects in BWR 8x8 Fuel Design and Performance", Reactor Safety Branch, DOR, May 1978. C-l/C-2
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