ML20069J160
| ML20069J160 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/31/1982 |
| From: | Charnley J, Hilf C, Zarbis W GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18025B889 | List: |
| References | |
| DRF-L12-00306-1, DRF-L12-306-1, Y1003501A40, NUDOCS 8210250074 | |
| Download: ML20069J160 (36) | |
Text
... . --- -
SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY
- NUCLEAR PLANT UNIT 2, RELOAD NO. 4 (CYCLE 5)
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Y1003J01A40 DRF L12-00306-1 Rev. O Class I July 1982 I
, /
l SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR
[
BROWNS FERRY NUCLEAR PLANT (CYCLE 5)
UNIT 2, RELOAD No. 4 l
l Prepared: !
C. L. Hilf Verified:
W. A. Za b s Approved:
J. S. Charnley, Manage Reload Fuel Licensing i
NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE, CAllFORNIA 95125
) GENERAL $ ELECTRIC i
Y1003J01A40 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPCRT PLEASE REAJ CAREFULLY This report vas prepared by General Electric solely for The Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory Conmission (USNRC) for amending TVA 's opemting license of the Brouns Ferry Nuclear Plant Unit 2. The information contained in this report is believed by General Electric to be an accurate and true represen-tation of the facts knoun, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting infomation in this document are contained in the contract betueen The Tennessee Valley Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Browns Ferry Nuclear Ptant Units 1 and 2, dated June 17, 1966, and nothing contained in this document shali be construed as changing said con-tract. The use of this information except as defined by said con-tract, or for any purpose other than that for chich it is intended, is not authorised; and uith respect to any such unauthorised use, neither General Electric Company nor any of the contributors to this document makes any representation or uarranty (express or implied) as to the con;pleteness, acciancy or usefulness of the information contained in this document or that such use of such information may not infringe privately cuned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
11
7 l
l Y1003J01A40 Rev. 0
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- 1. PLANT UNIQUE ITEMS (1.0)*
i Data for Sections 4 and 5 provided by Tennessee Appendix A f Valley Authority (TVA) f t
Safety / Relief Valve Capacity Appendix B f
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel fype Cycle Loaded Number Number Drilled Irradiated 8DB274L 2 8 8 8DRB284L 3 232 232 8DB274L 3 36 36
- e. P8DRB284L 4 240 240 i New P8DRB284L 5 168 168 P8DRB265H 5 80 80
[
Total 764 764
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at 19155 mwd /ST end of cycle:
Minimum previous cycle core average exposure at 18755 mwd /ST end of cycle from cold shutdown considerations:
Assumed reload cycle core average exposure at 18235 mwd /ST end of cycle:
Core loading pattern:
F5m1
- ( ) refers to area of discussion in " General Electric Standard Application for Reactor Fuel", NEDE-240ll-P-A-4, January 1982.
1 l
\
Y1003J01A40 Rev. 0
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)
See Appendix A
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
See Appendix A
- 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2)
(REDY EVENTS ONLY)
EOC 5 Void Fraction (%) 39.8 Average Fuel Temperature (*F) 1318 Void Coef ficient N/A* (c/% Rg) -6.85/-8.56 Doppler Coefficient N/A (c/*F) -0.224/-0.213 Scram Worth N/A ($) -46.31/-37.05
- 7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2) a ing a rs Fuel Bundle Power Bundle Flow Initial Design (Local Radial Axial) R-Factor (MWt) (1000 lb/hr) MCPR BOC 5 to EOC 5 P8x8R 1.20 1.51 1.40 1.051 6.352 108.0 1.27 8x8R 1.20 1.57 1.40 1.051 6.619 105.6 1.23 8x8 1.22 1.39 1.40 1.098 5.855 106.9 1.25
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization : No Recirculation Pump Trip : Yes Rod Withdrawal Limiter : No Thermal Power Monitor : Yes**
Measured Scram Time : No Number of Exposure Points : 1
- N = Nuclear input data A = Used in transient analysis
- No credit for the thermal power monitors was used in the analysis.
2
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- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) k Flux Q/A Transient (%NBR) (%NBR) P8x8R 8x8R 8x8 Figure Exposure: BOC 5 to EOC 5 599 122 0.21 0.18 0.18 2 Load Rejection w/o Bypass 4
Exposure: BOC 5 to EOC 5 122 122 0.13 0.13 0.12 3 Loss of Feedwater Heater Exposure: BOC 5 to EOC 5 385 120 0.16 0.15 0.14 4 Feedwater Controller Failure
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(S.2.2.1)
Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: Yes L
(
Rod Block Rod Position Reading (feet withdrawn) P8x8R/8x8R* 8x8R/P8x8R 104 3.5 0.10 14.5 105 4.0 0.12 15.0 106 4.0 0.12 15.0 107 4.5 0.13 15.2 108 5.0 0.14 15.2 109 5.5 0.15 15.2 110 6.0 0.16 15.2 Set point selected is: 106 l
- The 8x8 fuel type is not limiting since it is highly-exp*ed, low-reactivity fuel located primarily on the periphery of the core and not adjacent to any control blades whose worth is near that of the error rod.
3
Y1003J01A40 Rev. 0
- 11. CYCLE MCPR VALUES (S.2)
Nonpressurization Events Exposure Range: BOC 5 to EOC 5 P8x8R 8x8R 8x8 Loss of Feedwater Heating 1.20 1.20 1.19 Fuel Loading Error 1.22 Rod Withdrawal Error 1.19 1.19 Pressurization Events:
Exposure Range: BOC 5 to EOC 5 Option A Option B P8x8R 8x8R 8x8 P8x8R 8x8R 8x8 Load Rejection Without Bypass 1.34 1.30 1.30 1.24 1.22 1.22 Feedwater Controller Failure 1.28 1.27 1.26 1.25 1.24 1.23
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3) sl v (psig) (psig) Plant Response Transient __
MSlV Closure 1218 1254 Figure 6 (Flux Scram)
- 13. STABILITY ANALYSIS RESULTS (S.2.4)
Rod Line Analyzed: 105%
Decay Ratio: Figure 7 Reactor Core Stability Decay Ratio, x2 /*0 0.74 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 Channel Type 8x8R/P8x8R 0.29 8x8 0.38 4
t Y1003J01A40 Rev. O
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- 14. LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial MCPR Resulting MCPR l Misoriented 1.20 1.07
- 15. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
Bounding Analysis Results:
Doppler Reactivity Coefficient: Figure 8 Accident Reactivity Shape-Functions: Figures 9 and 10 Scram Reactivity Functions: Figures 11 and 12 Plant Specific Analysis Results:
Parameter (s) Not Bounded, Cold: Accident Reactivity Resultant Peak Enthalpy, Cold: 264.5 cal /gm Parameter (s) Not Bounded, HSB: None Resultant Peak Enthalpy, HSB:
- 16. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)
Refer to " Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 2", General Electric Company, February 1978 (NEDO-24088-1, as amended).
5
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Y1003J01A40 Rev. 0 2 6 10 14 18 22 26 30 .
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47 40 36 36 43 10 2 6 8 39 36 40 44 35 10 6 '8 0 31 40 36 36 44 NOTES: 1. Rod pattern is 1/4-core mirror symmetric.
- 2. Numbers indicate number of notches withdrawn out of 48.
Blank is a withdrawn red.
- 3. Error rod is (30, 35).
I Figure 5. Limiting RWE Rod Pattern
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Y1003J01A40 Rev. O AF ATURAL C: RCULATIO 1 B1 05 PERCENT ROD LI 4E CL LTIMATE STABILITY LINE 1.00 C C A
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O. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 7. Reactor Core Decay Ratio 12
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Figure 8. Doppler Reactivity Coefficient Comparison for RDA 13
Y1003J01A40 Rev. 0 20.O A ACCIDENT FUNCTION 8 BOUNDING VALU E 280 CAL /G I
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- 0. 0 5.0 10.J 15.0 20.0 ROD POSITION, FEET OUT Figure 9. Accident Reactivity Shape Function Cold Startup 14
l Y1003J01A40 Rev. 0 20.O A ACCIDENT FUNCTION 8 BOUNDING VALU E 280 CAL /G t
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Y1003J01A40 Rev. 0 50.O A SCRAM FL NCTION B BOUNDING VALUE 280 CAL /G 40.0 m
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16
Y1003J01A40 Rev. 0 70.O A SCRAM FL NCTION 8 BOUNDING VALUE 280 CAL /G 60.0 w
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1.0 2.0 3.0 4.0 5. 0 6.0 ELAPSED TIME, SECONDS Figure 12. Scram Reactivity Function Hot Startup 17/18
Y1003J01A40 Rev. O APPENDIX A SHUTDOWN MARGIN DETERMINATION A.1 BASES The reference loading pattern, documented in Item 3 of this supplemental reload submittal, is the basis for all reload licensing and operational planning and is comprised of the fuel bundles designated in Item 2 of this supplemental submittal. It, in turn, is based on the best possible predic-tion of the core condition at the end of the present cycle and on the desired core energy capability for the reload cycle. It is designed with the intent that it will represent, as closely as possible, the actual core loading pattern.
A.2 CORE CHARACTERISTICS The reference core is analyzed in detail to ensure that adequate shutdown margin exists. This section discusses the results of core calculations for shutdown margin (including the liquid poison system).
A.2.1 Core Effective Multiplication and Control Rod Worth Core effective multiplication and control rod worths were calculated using the TVA BWR simulator code (Reference A-1, A-3) in conjunction with the TVA lattice physics data generation code (References A-2, A-3) to determine the core reactivity with all rods withdrawn and with all rods inserted. A tabulation of the results is provided in Table A-1. These three eigenvalues (effective multiplication of the core, uncontrolled, fully controlled, and with the strongest rod out) were calculated at the beginning-of-cycle 5 core average exposure corresponding to the minimum expected end-of-cycle 4 core average exposure. The core was assumed to be in a xenon-free condition.
Cold k,gf was calculated with the strongest control rod out at various exposures through the cycle. The value R is the difference between the strongest rod out k at B0C and the maximum calculated strongest rod out ff 19
Y1003J01A40 Rev. O k
eff at any exposure point. The strongest rod out keff at any exposure point is equal to or less than:
SRO k = (Fully Controlled keff)BOC + (Strongest Rod Worth)BOC + R eff A.2.2 Reactor Shutdown Margin Technical Specifications require that the refueled core must be capable of being made suberitical with 0.38% Ak margin in the most reactive condition throughout the subsequent operating cycle with the most reactive control rod-in its full out position and all other rods fully inserted. The shutdown margin is determined by using the BWR simulator code to calculate the core multiplication at selected exposure points with the strongest rod fully with-drawn. The shutdown margin for the reloaded core is obtained by subtracting the k e
given in Table A-1 from the critical keff f 1.0, resulting in a calculated cold shutdown margin of 1.4% Ak.
A.2.3 Standby Liquid Control System The standby liquid control system (SLCS) is designed to provide the capability of bringing the reactor, at any time in a cycle, from a full power and minimum control rod inventory (which is defined to be at the peak of the xenon tran -
sient) to a suberitical condition with the reactor in the most reactive xenon-free state.
The SLCS shutdown margin is determined by using the BWR simulator code to calculate the core multiplication for the cold, xenon-free, all rods out condi-tions at the exposure point of maximum cold reactivity, with the soluble boron cocentration given in the technical specifications. The resulting k-effective is subtracted from the critical k-effective of 1.0 to obtain the SLCS shutdown margin. Table A-2 gives the results of the SLCS evaluation.
20
Y1003J01A40 Rev. O Table A-1 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL ROD WORTilS - NO VOIDS, NO XENON, 20*C Uncontrolled, k 1.115 Fully Controlled, k 0.955 Strongest Control Rod Out, k e 0.986 R, Maximum Increase in Cold Core 0.000 Reactivity With Exposure Into Cycle, Ak Table A-2 STANDBY LIQUID CONTROL SYSTEM CAPABILITY Shutdown Margin (Ak) m (20*C, Xenon Free) 600 0.023 a
21 m
. . . l l
Y1003J01A40 Rev. O References i A-1. S. L. Forkner, G. H. Meriwether, and T. D. Beu, "Three-Dimensional LWR Core Simulation Methods", TVA-TR78-03A, 1978.
A-2. B. L. Darnell, T. D. Beu, and G. W. Perry, " Methods for the Lattice Physics Analysis of BWRs", TVA-TR78-02A, 1978.
i A-3. " Verification of TVA Steady-State BWR Physics Methods", TVA-TR79-01A, 1979.
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Y1003J01A40 Rev. O APPENDIX B SAFETY / RELIEF VALVE CAPACITY AT SET POINT (NO./%): 12/77.6*
- Assumed one safety / relief valve out of service.
Reference pressure was 1105 + 1% psig.
23/24 (FINAL)
_ _ _ _ _ .