ML19344E654
| ML19344E654 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/31/1980 |
| From: | Engel R, Hilf C GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18025B055 | List: |
| References | |
| Y1003J01A03, Y1003J1A3, NUDOCS 8009020412 | |
| Download: ML19344E654 (29) | |
Text
{{#Wiki_filter:- . Y1003J01A03 g Class I W August 1980 f I
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SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY NUCLEAR POWER STATION UNIT 3 RELOAD NO. 3
)
l I l Prepared: d.< / C. L. Hilf Approved: ' ' .- R. E. Engel, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 95125 G EN ER AL $ ELi!CTRIC c Boo:o10 eel)
' Y1003J01A03 1
1 IMPORTANT NOTICE REGARI,ING CONTENTS OF THIS REPOET 1 f 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 l Unit 3. The information. contained in-this report is believed by General l- Electric to be an accurate and true' representation of the facts known, j' obtained or provided to General Electric at the time this report was prepared.
l-The only undertakings of the General Electric Company respecting information in this document are contained in the contract between The Tennessee Valley l-
' Authority and General Electric Company for nuclear fuel and related services for the-nuclear system for Browns Ferry Nuclear Plant Unit 3, dated June 17, 1966, and nothing contained in this document shall be constructed as changing said contract. The use of this information except as defined by said contract, I
or for any purposes other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric , I j 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 information contained in this document or that such use of such infor-mation may not infringe privately owned rights; nor do they assume any respon-
.sibility for liability or damage of any kind which may result from such use of such information.
l f S s
l l , Y1003J01A03
- 1. PLANT-UNIQUE ITEMS (1.0)*
7 8 Items different from or not included in Reference 1:
/
Data for Section 4 provided by Tennessee Valley Authority (TVA): Appendix A Fuel Loading Error LHGR: Appendix B Safety / Relief Valve Capacity: Appendix B Spring Safety Valve Capacity: Appendix B Rated Steam Flow: Appendix B GETAB Analysis Initial Conditions: Appendix B New Bundle Loading Error Event Analysis Procedures: Reference 3 Margin to Spring Safety Valves: Appendix C
- 2. RELOAD FUEL BUNDLES (1.0. 3.3.1 and 4.0)
I Fuel Type Number Number Drilled Initial Core SDBil9 288 288 Reload 1 8DRB265L 208 208 Reload 2 P8DRB265L 144 144 New P8DRB265L 124 124 TOTAL 764 764
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycle: 14,297 mwd /t Assumed reload cycle core average exposure at end of cycle: 15,105 mwd /t Core loading pattern: Figure 1
*( ) Refers to areas of discussion in Reference 1.
1
~
l 1
-Y1003J01A03-i .
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CORE SYSTEM f WORTH - NO VOIDS. 20*C (3.3.2.1.1 AND 3. 3. 2.1. 2)
See Appendix A for this data provided by The Tennessee Valley Authority. , 5 .' STANDBY LIQUID CONTROL' SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) l i Shutdown Margin (Ak) ! pjgt (20*C, Xenon Free) l 600 0.04
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)
Void Coefficient N/A* (c/% Rg) -6.97/-8.71 Void Fraction (%) 40.29 Doppler Coefficient N/A (c/*F) -0.228/-0.217 l Average Fuel Temperature (*F) 1343 Scram Worth N/A ($) -37.67/-30.13 Scram Reactivity vs Time Figure 2
-7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) l 8x8 8x8R P8x8R Exposure- EOC 4 EOC 4 EOC 4 Peaking factors (local, radial ~ 1.22 1.20 1.20 and axial) 1.42 1.55 '1.55 1.40 1.40~ 1.40 R-Factor 1.098 1.051 1.051 Bundle Power (MWt) 5.987 6.550 6.526 -Bundle Flow (103 lb/hr) 108.2 108.5 109.2 Initial MCPR 1.24 1.25 1.25 i *N = Nuclear Input Data A = Used in Transient Analysis 2
Y1003J01A03
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
Recirculation Pump Trip
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.2)
F Core Power Flow 4 Q/A 's! v ACPR Plant Transient Exposure (%) (%) (% NRR) (? NBR) (psig) (psig) 8x8 8x8R P8x8R Response Load Rejection BOC4-EOC4 104.5 100 239 111 1226* 1250 0.17 0.18 0.18 Figure 3 Without Bypass loss of 100*F -- 104.5 100 128 123 1013 1069 0.15 0.15 0.15 Figure 4 Feedwater Heating Feedwater BOC4-EOC4 104.5 100 164 112 1155 0.12 1189 0.12 0.12 Figure 5 Controller Failure
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
TRANSIENT
SUMMARY
(5.2.1) ACP R*
- MLHGR 8 **
Rod Block Rod Position 8x8R/ 8x8R/ Limiting Reading (Feet Withdrawn) 8x8 P8x8R 8x8 P8x8R Rod Pattern 104 3.5 0.10 0.10 14.8 16.2 Figure 6 105 4.0 0.12 0.11 15.2 16.5 Figure 6 i 106* 4.5 0.14 0.12 15.2 16.6 Figure 6 , 107 4.5 0.14 0.12 15.2 16.6 Figure 6 108 5.5' O.18 0.14 15.2 16.7 Figere 6 109 6.5 0.20 0.16 15.2 16.7 Figure 6 110 7.5 0.20 0.17 1 .2 16.7 Figure 6
- Indicates setpoint selected.
**The initial MCPR (1.24) for the 8x8R and P8x8R fuel was 0.01 less than the
. operating limit MCPR (1.25). This is discussed on pp. B-114 and B-115 of Reference 1.
***A 2.2% peaking penalty for densification is included.
+Less than 25 psi margin to spring safety valves. One safety / relief valve is assumed out of service. See Appendices B and C.
3
= -
' Y1003J01A33 '
l 1 l
- 11. OPERATING MCPR LIMIT (5.2)
BOC4 to EOC4 1.24 8x8 Fuel 1.25 8x8R Fuel 1.25 P8x8R Fuel
- 12. OVERPRESSURIZATION ANALYSIS SUMHARY (5.3) f Power . Core Flow 'sl y Plant
) Transient (%) (%) (psig) (psig) Response MSIV Closure 104.5 100 1265 1299 Figure 7 , (Flux Scram) i
- 13. STABILITY ANALYSIS RESULTS (5.4) 4 Decay Ratio: Figure 8 Reactor-Core Stability:
Decay Ratio,.x2 /*0 0.85 (105% Rod Line - Natural Circulation Power) j Channel Hydrodynamic Performance ' ) Decay Ratio, x2 /*0 (105% Rod Line - Natural Circulation Power) j 8x8R/P8x8R 0.29 8x8 0.36 l
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
[' Reference 2. i e 4
Y1003J01A03
- 15. LOADING' ERROR RESULTS (5.5.4)
Limiting Event: Rotated Bundle P8DRB265L MCPR: 1.C8 i
- 16. CONTROL ROD' DROP ANALYSIS RESULTS (5.5.1) 1 Doppler Reactivity Coefficient: Figure 9
. Accident Reactivity Shape Functions: Figures 10 and 11
", Scram Reactivity Functions: Figures 12 and 13 Plant specific analysis results ] Parameter not bounded: None t k 4 i i 4 e E i i 5
Y1003J01A03 - ; i REFERENCES 'l. General Electric Boiling Water Generic Reload , Fuel Application, NEDE-240ll-P-A, August 1979.
- 2. -Loss-Of-Coolant Accident Analysis Report for Browns Ferry Nuclear Plant Unit 3, NEDO-24194A, July 1979.
- 3. Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 3 Reload 1, NEDO-24128 (Appendix A), June 1978.
)
6
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EMMMMMMMMME r " M M M M M BBBE" lIIIIIIIIIIIII I 1 3 5 7 911131517192123252729313335373941434547495153555759 FUEL TYPE : A = 808219 ,1C' E = P80R8265L,R3 8 = 808219 ,1C F = 808219 .lc C = 8DR8265L ,R1 G = 808219 ,1C D = P80R8265L,R2 H = 80R8265L ,R1 'lC = INITIAL COR E. R1 = RELO AD 1, R2 = R ELOAD 2, etc. Figure 1. Reference Core Loading Pattern 7
Y1003J01A03
.100 " 45 C-679 CR0 IN PERCENT 1-NOMINAL SCRAM CURVE IN (-51 90- 2-SCRAM CURVE USED IN ANALYSIS
-40 80-
-35 70-
' 30 E G EI 60- 1 5 >-
S -25 t z 3 2 50- [i 2 5 8 o_
-20 40-
-15 30- l 1
-10 1 20-1 10- -5 0 L~ , , , O l 0 1 2. 3 4 TIME (SECCN05)
Figure 2. Scram Reactivity and Control Rod Drive Specifications 8
1' 1 NEUTRON FLUX 1 VESSEL PFES RISE (PSil 2 AVE SURFFCE HEAT FLUX 2 SAFETY VfLVE FLOW 150. 3 CORE INLE T FLOW g 3 F L OW 6
~ 100 200' '
!. \\ '
b 50 ' V-
- 100 N 3 N ~1 h
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- 0. 2 *- 2u 2' 2A *
- 4. 8. 12. 16. O. 4. 8. 12. 16.
TIE (SEC) TIM (SECl d 8
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l LEVEL (INCH-N F-SEP-SKIRT 1 VOIO REA TIVITT 5 o 2 VESSEL $1EAMFLOH 1 2 ^,G^~ C^ g C;;C' A Q w 200' 3 TUR8INE 5 TEAMFLOW 4 t LLUWHi tt FLOW 3' [ 3 SCRAM RffCTIVITY 4 IDIHL IUCTIVITT g 100' '*
- 5 0' 84
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- u f 6 a.
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-100. i-0.
4
- 2. - i- - *
- 4. 8. 12. 16. O. 0.4 0.8 1.2 1.6 TIME ISECl TIME (SEC)
Figure 3. Plant Response to Generator Load Rejection Without Bypass
1 WUTntyi FLL; I YESSEL PFES RISE (PSil u 3 0VC- LiencE-KAT-RUX 2 TLIEF VfLVE FLOW 3 BTPASS VILVE FLOW ISO-g' 3 CnnF_Ita.1 TJU1H
;ccmiun suu 125-0 1* 1>
3 3 3 3 75. 8 100. - M IE b e-5 S0. 2s. 1 I I
" c3
-tm / -13 11
- 0. *- * -25.
- 0. 40. 80. 120. 160. O. 40. 80. 120. 160.
TIME ISEC) TIME (SEC) 4 o s -
@o 1 LEVEL (IMH-EF-SEP-SKIRT I V010 EMTIVITT 2 VESSEL S1EAMFLOW 2 00PPLER TEACTIVITY 3 SCRAM REfCTIVITY O
3 itEUINE
- TEAMFLOW 3*
150* ~ 4 FELIl4ATEFTLCN -- DOTWcTIVIIT 5 . 1 1 A O. 100. 2 2 su
$0. .
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-1 1 1 1 R
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* -2. *-
- 0.~ '- -
- 0. 40. 80. 120. 160. O. 40. 80. 120. 160.
TIME ISEC) TIME (SEC) Figure 4. Plant Response to Loss of 100 Deg F Feedwater lleating
4 150' \ ' "'"'" " "* 2 AVE SURFFCE HEAT FLUX MCM 4 CORE INLt i SUB I ILW 125'
'[\
' \
2 SAFETT VFLVE FLOW 3 RELIEF VFLVE FLOW 4 BTPRSS VfLVE FLF
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- 25. .
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- 0. *- ' 1 -
- 25. *- -
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- 10. 20. 30. 40. O. 10. 20. 30. 40.
Tiff (SEC) TIME (SEC1 w - 8 W 0 o 1 LEVELIINCH-REF-SEP-SKIRT ! VOID REAC TIVIT o 2 VESSEL S1EAMFLOM 2 SOPPLER FEAC ITY W 150 - 3 TUR8INE 5 TEAMFLOp_ g* M CRAM REFCT ITT 4 tuUNHlu FLN # IFL fEFETIVIIT a u 5 f I M M b I l
- I 100' 4
1 M z
- 0. 26'- '
>)
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_ Ny b L e ' 50' '
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- O -1"
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- 2. - '- - ll *
- 0. 10. 20. 30. 40. O. 6. 12. 18. 24.
TIME (SEC) TIME iSECl Figure 5. Plant Response to Feedwater C mtroller Failure, Maximum Demand
Y1003J01A03 . NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SY;2iETRIC (FULL CORE SHOWN) l 2. NO. INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD
- 3. ERROR ROD IS~(26, 35) 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 6 4 4 6 53 36 36 36 36 36 l
51' 6 6 2 2 6 6 47 36 36 36 36 36 36 36 l 43 6 6 10 14 14 10 6 6 39 16 36 36 40 36 36 36 35 4 2 14 0 0 14 ~2 4 31 36 36 40 36 40 36 36 27 4 2 14 0 0 14 2 4 l 23 36 36 36 40 36 36 36 l 19 6 6 . 10 14 14 10 6 6 15 36 36 36 36 36 36 36~ 11 6 6 2 2 6 6 7 16 36 36 36 36 l ! 3 6 4 4 6 Figure 6. Limiting RWE Rod Pattern 12 _ . ~_ - . __
1 NEUTRON FLUX ! VESSEL PFES RISE IPSI) I 2 AVE SufffCE HEAT FLUX 2 SAFETT VFLVE FLOW
~
'SO*
* - 3 CORE INLE T FLOW g* 3 BE_tJEF VFLVE FLOW 4 4 BIPRSS VFLVE FLOW 5 5 N3 6 n 100, 200. !
is k
\\' -
b SO ' - 100 b 1 3 3
' 3
- 0. - '- l ' 1 ' '. - u --- nu zu O. 'l .
O. 4. 8. 12. 16. O. 4. 8. 12. 16. TIME (SECl TIME (SEC) - - 8 u W o 1 LEVEL (INCH-E F-SEP-SKIRT I VOIO ERCTIVIT C 2 VESSEL S1EAMFLOW 2 00PPLER 1 v!TT W 200* 3 TUR8INE 5 IEAMFLOW y* 3 SCRAPiAGCTIVITY 4 FEEDWATEF FLOW 4 IDI HEFCTIVITT S E 100. h 0. ' ' 3 2 N g
.w ,, 7 0.
A b# fVH 3 1 l t . 1.
.t g -
M 100. '- * -2. - t- *
- 0. 4, 8. 12. 16. O. 0.6 1.2 1.8 2.4 TIME (SEC) TIME (SEC1 Figure 7. Plant Response to MSIV Closure
Y1003J01A03 1.2 ULTIMATE STABILITY LIMIT 1.0 = = = = = - - = = = - - = = = = = = = = - = = = - - - - - - = = - - E w 0.8 - X NATURAL 9 CIFICULATION E e l 8 0.s - 106% ROD LINE O.4 0.2 - 0 ! ! ! ! I O 20 40 60 80 100 PERCENT POWER Figure 6. Decay Ratio 14
Y1003J01A03 0 ' A CALCULATED VALUE-COLD 5 CALCULATED VALUE.HSB C BOUNDING VALUE FOR 280 cal /g, COLD D BOUNDING V ALUE FOR 280 cal /g, HSB i
~5
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V
-{-'
d_ E
-15 u J
N E c -20 Y e 8 5 i a- - 25 8-
/
< -30
-35 0 500 .1000 1500 2000 2500 3000 FUEL TEMPERATURE (dog C)
Figure 9. Doppler Reactivity Coefficient Comparison for RDA 15
-t 7 ,
1 Y1003J01A03 20 l A ACCIDENT FUNCTION 8 80UNO,1NG,VALUE FO,R,290 cmi/s 16 E h
@ a a .__ a g rg*~
- 2 e
i w
$ 10
/
e. b t 5 E N e 1
$ fl 1
i 1 0, , 0 5 10 15 20 ROD POSITION, feet OUT Figure 10. Accident Reactivity Shape Function at 20*C 16
'.Y1003J01A03 i
A ACCIDENT FUNCTION 8 BOUNDING VALUE FOR 280 col /g 15 k n_n 8 F" e e w g 10 o. b 5 b 6 e 5 u i 0 5 5 10 15 20 ROD POSITION, feet OUT Figure 11. Accident Reactivity Shape Function at 286*C 17
Y1003J01A03 , . l l 50 i A SCRAM FUNCTION 8 BOUNDING VALUE FOR 280 cal /g 40 ' G' I O z 8 3 O ! I 30 l D ! M 4 i t b m
- O i
6 w ! Z
- ~. 20 b
2 U w ! E l l l l_ i l 0 7 d 1 2 3 4 5 6 ELAPSED TIME, seconds Fi l ;ure 12. Scram Reactivity Function at 20*C , 18 m,
Y1003J01A03 4 I I A SCRAM FUNCTION 75 8 BOUNDING VALUE FOR 280 cel/g G
?
O 2 I 8 50 I b. x h - a 0 3 m E N s E o 25 L g . e 0 - l 6 1 3 3 4 5 6 ELAPSED TIME, seconds Figure 13. Scram Reactivity Function at 286*C 19/?0
Y1003J01A03 APPENDIX A SHUTDOWN MARGIN DETERMINATION A.l~ 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 prediction 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 cold shutdown margin exists. This section discusses the results of core calculations for shutdown margin.
- 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 (References 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 multi-plication of the core, uncontrolled, fully controlled, and with the strongest rod out) were calculated at the beginning-of-cycle 4 core average exposure corresponding to the minimum expected end-of-cycle 3 core average exposure.
The core was assumed to be in a xenon-free condition. m A-1 _ -n- w t
Y1003J01A03 Cold k,gg 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,gg at BOC and the maximum criculated strongest rod out k,gg at any exposure point. The strongest rod out k,gg at any exposure point is equal to or less than: S k,RO gg = (Fully controlled k gg) BOC + (Strongest Rod Worth) BOC + R J A.2.2 Reactor Shutdown Margin Technical Specifications require that the refueled core must be capable of being made suberitical with 0.38 percent 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 calcu-late the core multiplication at selected exposure points with the strongest rod fully withdrawn. The shutdown margin for the reloaded core is obtained by subtracting the k, 0 given in Table A.1 from the critical k,gg of 1.0, resulting in a calculated cold shutdown margin of 1.1 percent Ak. A-2
Y1003J01A03 Table A.1 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL
. ROD WRTHS - NO VOIDS, NO XENON, 20*C Uncontrolled, K 1.120 Fully Controlled, K.
0.955 Strongest Control Rod Out, K, 0.989 i
- R, Maximum Increase in Cold Core Reactivity 0.000 With Exposure Into Cycle, ok l
9 e i i . A-3'
I Y1003J01A03 REFERENCES f A-1. S.~ L. Forkner, G. H.-Meriwether, and T. D. Beu, "Three-Dimensional LWR Core Simulation Methods," TVA-TR78-03A, 1978
' A< ' . B.~L. Darnell, T. D. Beu, and G. W. Perry, " Methods for the Lattice Physics Analysis of LWR's," TVA-TR78-02A, 1978 I A-3. " Verification of TVA Steady-State BWR Physics Methods," TVA-TR79-01A, j 1979 l
l l l i l i i l 4 t A-4
Y1003J01A03 APPENDIX B Fuel Loading Error LHGR*: Rotated Bundle, 16.9 kW/ft; Misplaced Bundle, 18.1 kW/ft Safety / Relief Valve Capacity at Setpoint (No./%): 10/63.6** Spring Safety Valve Capacity at Setpoint (No./%): 2/14.2 6 Rated Steam Flow: 14.09 x 10 lb/hr GETAB Analysis Initial Conditions Reactor Pressure: 1035 psia Inlet Enthalpy: 521.5 Btu /lb e
*2.2% peaking penalty for densification is included.
** Assumes one safety / relief vlave out of service.
B-1/B-2
Y1003J01A03 APPENDIX C MARGIN TO SPRING SAFETY VALVES f The rationale for changing the basis for providing pressure margin to the spring safety valves is presented in Reference C-1. This change has been-
- accepted by the NRC (Reference C-2).
On this basis the plant can operate at full power throughout the cycle. i i d A
=
i
^
C-1
Y1003J01A03 . REFERENCES C-1. J. F. Quirk (GE) letter to Olan D. Parr (NRC), " General Electric Licensing Topical Report NEDE-240ll-P-A, . ' Generic Reload Fuel Application, ' Appendix D, Second Submittal," dated February 28, 1979. C-2. Letter, T. A. Ippolito (NRC) to D. L. Peoples (Commonwealth Edison Co.) enclosing a Safety . Evaluation supporting Amendment No. 42 to Facility 4 Operating License No. DPR-25 Dresden Nuclear Power Station Unit 3, dated April 16, 1980. 1 l 4 C-2}}