ML20077K929

Rev 0 to Supplemental Reload Licensing Submittal for Browns Ferry,Unit 1,Reload 5

ML20077K929
Person / Time
Site:	Browns Ferry
Issue date:	05/31/1983
From:	Charnley J, Elliott P, Hilf C GENERAL ELECTRIC CO.
To:
Shared Package
ML18025B986	List: ML20076L991 ML20077K929
References
22A8559, 22A8559-R, 22A8559-R00, NUDOCS 8307200105
Download: ML20077K929 (30)

Text

_ _ _ _ _ _

i MAY 1983 l

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY UNIT 1 RELOAD 5 930720u10b 830713 DR ADOCK OS000

22A8559 Revision 0 Class I May 1983 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY UNIT 1, RELOAD 5 Prepared:

7 . Elliott

/

Verified:

C. L.lIllf Approved: A J//S. Charnley, Prograpnager Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION

• GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 95125 GENER AL h ELECTRIC i

22A8559 Rev. 0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for 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 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 contract between Tennessee Valley Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Browns Ferry 1 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 such unauthorized use, neither General Electric 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 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.

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22A8559 R:v. 0

1. PLANT UNIQUE ITEMS (1.0)*

A. Information for Sections 4 and 5 Provided by the Appendix A Tennessee Valley' Authority B. Plant Parameter Differences Appendix B C. Increased Core Flow, 105% Rated Appendix C

2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number Number Drilled Irradiated 8DB274L 2 17 17 8DRB265H 3 4 4 P8DRB284L 4 231 231 P8DRB284L 5 220 220 P8DRB265L 5 36 36 GLTA-l** 5 2 2 GLTA-2** 5 2 2 New P8DRB284L 6 44 44 6 8 8 P8DRB284Z P8DRB265H 6 164 164 P8DRB284L 6 36 36 Total 764 764

• ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel". NEDE-240ll-P-A (latest approved revision); a letter "S" preceding the number refers to the appropriate country-specific supplement.

• • Previously described in " Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Power Plant Unit 1 Reload No. 4 (Cycle 5),"-

Y1003J01A19. Rev. 1 (Appendix E), September 1982.

1

22A85591 R2v. 0

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominalpreviouscyclecoreaverage.:epposureadend ,

of cycle: -

19074 mwd /St Minimum previous cycle core average exposurefat end of cycle from cold shutdown considerations: 18751 mwd /St Assumed reload cycle core average exposure at end of cycle: 18045 mwd /St Core loading pattern: Figure"3c. ,

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO ~

VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2) '

P See Appendix A A

5. STANDBY LIQUID CONTROL SYSTEM SHUTD0'WN CAPABILITY. (3.3.2.1.3) t See Appendix A

6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT- (3.3.2.1.5 AND S.2.2) ,

(REDY Eventq Only)

EOC 6 ,

Void Fraction (%) 39.2 Average Fuel Temperature (*F) 1273 Void Coefficient N/A* (c/% Rg) -7.10/-8.88 Doppler Coefficient N/A* (c/*F) -0.228/-0.217 Scram Worth **

6

• N = Nuclear Input Data; A = Used in Transient Analysis

• • Generic, exposure independent values are used as given in " General Electric Boiling Water Reactor Generic Reload Fuel Application", NEDE-24011-P-A-1, Amendment 10, April 1981.

2

__ . . . . . a

22A8559 Rsv. 0

7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) en ng ac rs Bundle Power Bundle Flow Initial Fuel Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR BOC 6 to EOC 6 P8x8R 1.20 1.52 1.40 1.051 6.411 114.6 1.28 8x8R 1.20 1.55 1.40 1.051 6.512 112.5 1.25 8x8 1.22 1.41 1.40 1.098 5.947 112.3 1.24

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.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

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single Loop Operation: Yes Load Line Limit: No Extended Load Line Limit: No Increased Core Flow: Yes Flow Point Analyzed: 105%

Feedwater Temperature Reduction: No

• No credit _ for the thermal power monitor-was used in the analysis.

3

I 22A8559 Rev. 0

10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Flux Q/A Transient (% NBR) (% NBR) P8x8R 8x8R 8x8 Figure Exposure: BOC 6 to E0C 6 611 123 0.21 0.18 0.17 2 Load Rejection Without Bypass Exposure: BOC to EOC 123 123 0.14 0.14 0.14 3 Loss of Feedwater Heater Exposure: BOC 6 to E0C 6 398 121 0.16 0.15 0.14 4 Feedwater Controller Failure

11. LOCAL ROD WITRDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(S.2.2.1)

(Generic Bounding Analysis Results)

Rod Block ACPR Reading (all fuel types) 104 0.13 105 0.16 106 0.19 107 0.22 108 0.28 109 0.32 110 0.36 Setpoint Selected: 106

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events Exposure Range: BOC to EOC P8x8R 8x8R 8x8 Loss of Feedwater Heater 1.21 1.21 1.21 Fuel Loading Error 1.25 Rod Withdrawal Error 1.26 1.26 1.26 4

22A8559 Rev. O Pressurization Events Exposure Range: BOC 6 to EOC 6 Option A Option B P8x8R 8x8R 8x8 P8x8R 8x8R 8x8 Load Rejection Without Bypass 1.34 1.30 1.29 1.24 1.22 1.21 Feedwater Controller Failure 1.28 1.27 1.26 1.25 1.24 1.23

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) si v Transient (psig) (psig) Plant Response MSIV Closure 1220 1257 Figure 5 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: 105%

Decay Ratio: Figure 6 Reactor Core Stability Decay Ratio, x2 I*0 - 0.77 Channel Hydrodynamic Performance Decay Ratio, 2x /*0' Channel Type P8x8R/8x8R 0.29 8x8 0.38

{

y 15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial MCPR Resulting MCPR Misoriented 1.23 1.07 5

22A8559 Rev. 0

16. CONTROL R0D DROP ANALYSIS RESULTS (S.2.5.1)

Bounding Analysis Results:

Doppler Reactivity Coefficient: Figure 7 Accident Reactivity Shape Functions: Figures 8 and 9 Scram Reactivity Functions: Figures 10 and 11 Plant Specific Analysis Results:

Parameter (s) not Bounded, Cold: Accident Reactivity Scram Reactivity Resultant Peak Enthalpy, Cold: 167.2 Parameter (s) not Bounded, HSB: Accident Reactivity Resultant Peak Enthalpy,llSB: 243.7

17. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)

" Loss-of-Coolant Accident Analysis Report for Browns Ferry Unit 1",

General Electric Company, NEDO-24056, Rev. 1, May 1983.

l 6

22A8559 Rev. O LIST OF FIGURES

1. Reference Core Loading Pattern

2. Plant Response to Generator Load Rejection, Without Bypass

3. Plant Response to Loss of 100*F Feedwater Heating

4. Plant Response to Feedwater Controller Failure

5. Plant Response to MSIV Closure

6. Reactor Core Decay Ratio

7. Fuel Doppler Coefficient in 1/A*C

8. Accident Reactivity Shape Function, Cold Startup

9. Accident Reactivity Shape Function, Hot Startup

10. Scram Reactivity Function, Cold Startup

11. Scram Reactivity Function, Hot Startup 7

22A8559 Rev. 0

mMMMMMMMs

EMMMMMMMME:

esMMMMMMMMMMMos

sMMMMMMMMMMMMMs IMMMMMMMMMMMMMMM CMMMMMMMMMMMMMMM CMMMMMMMMMMMMMMM EMMMMMMMMMMMMMMM CHEMMMMMMMMMMMMM IMMMMMMMMMMMMMMM EMMMMMMMMMMMMMMM

*MMMMMMMMMMMMM" ll MMMMMMMMMMM"*

l EMMMMMMMMME

*MMMMMME" l IIIIIIIIIIIIi 1 3 5 7 911131517192123252729313335373941434547495153555759 FUEL TYPE B R 265H G=P RB284L

""it "

; " *:

E = P8DRB265L J = P8DRB284L Figure 1. Reference Core Loading Pattern 8

22A8559 Rev. 0 1 NEUTRCN FLUX X VESSEL PRESh RISE (PSI) 2 AVE SUFF ACE HEAT FLUX 2 SMETY VAL \t FLOW 3 CORE IfLET :LCW 3 FELIEF V ALb E FLOW 150.0 300.0 4 gvear vattg eggw 5 ,

y 100.0 b)y 1 200.3 m L s ~

x

"'0 100.0

/

0 0 0 0 0 0 I

1. • * - 0.0 , , , , _ , ,, , , ,

0. 3 2.0 4.0 6.0 0.0 2.0 4.0 6.0 TIME (SECONOS) TIPE (SECONDS) 1 LEVELt!NCH-2EF-SE P-SKR T l 1VO!DREACT!h!TY 2 VESSEL STE APFLOW 2 DOPPLER RE T!VITY 3 TURBihE STEAM'LCW 3, S,CR 200.0 i_rEEc-Agge r_gy 1.0 /\ gy ,A,M RE AC IVITY

_ e ac turvv b

I 200.0 .im 0. 0 ..

1M

~w 2

e i

N 4

aa J g-i.0 V

-100.0 -2.0

0. 0 2.0 4.0 6.0 0.0 2.0 4.0 6.0 TIME (SECONDS) TIME (SECONOS)

Figure 2. Plant Response to Generator Load Rejection, Without Bypass 9

22A8559 Rev. 0 1NELIRON FLUX 1 VES EEL PRrSS RISE (PSI) 2 AVE SUFF ACE HEAT FLUX

"" 2 FELIEF YALVE FLCW 3 C05 '- 4W 15 0. 9 W .'. _ Ee. 79 Y 3 BYP ASS VALVE FLOW

._g .. . .- , 100.0 O 100.0 $ " 3 3 3 0 3 3 3 3 i e

b 5 50.0 M

~ 50.0

0. 0 - - ' ' 2 s . , .- - - .- - . 7- --

• -1

-1

0. 0

0. 0 100.0 200.0 0. 0 100.0 200.0

!!ME (SECONDS)

IIME (SECONDS)

ILEV EL(INCH REF-SEP.SKET) 2 VESiEL STEAMFLOW

! VO! REACTIVITY 2 DCP LER REACTIVITY 3 TUR31NE STEAT LOW 3 SCR H REACT 1v!TY 150.0 tFEE M ten rity 1.0

' Tent cEact!r!Tv 4*=12= 's " MtW 3e; - 0 ,

j y

- [ - .  ::  :  :  :: : : : :,

S -2  :  :

?

50 8

?

V..0 W

0. 0

-2.0 0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECONDS 1 TIME (SECONOS)

Figure 3. Plant Response to Loss of 100*F Feedwater IIcating 10

22A8559 Rev. 0 150.0 1 NEUTkCta llk i VESSEL PR(? RISE (PSI) 2 AVE Sf (CE EAT FLUX 2 SAFETY VAVN FLCW 3 00EE I (T r W 3 REL IEF V A b FLOW 15 0. 0 erecr ruf7 ~L 4 BYPASS V A.V FLGW 100.0

~ ,. ,

T ' ~3 ' J 100.0 '

E i H

b 50.0 y

w U

t c 4

- 50.0 1

M =' a a ; 2  :

0. 0

0. 0 10.0 20.0 30.0 0. 0 10.0 20.0 30.0 TIME (SECONOS) TIME (SECONOS) 1 LEVEL (INCH-REF-SEP-SKAT)  ! VOID REACT!b!TY 2 VESSEL STE4fLGd 2 DOPPLER REACTIVITY SMILCW 3, SCR vnr A.M, c-RE, ACTIVITY

3. err N itra rTUABlhE nu STE 1.0 cr:urry 150.0

=

9'%

100.0 . 3 0.0 , . ;  ;; _ ,

2 S  !

j

, su t

4 l

=

=

50.0 ./ h. b h -1.0 i} lIf e

0.0 d , . -2.0 C. 0 ' 10.0 20.0 30.0 0. 0 10.0 20.0 30.0 TIME (SECONDS) TIME (SECONOS)

Figure 4. Plant Response to Feedwater Controller Failure 11

22A8559 Rev. 0 1 NEUTRON F UX 2 AVE SURFACE f fE AT FLUX 1 VESSEL PFESS R!SECPSI) 2 SAFETY VfLVE FLOW 3 CORE INLET FLOW 150.0 3 FELIEF V/LVE FLOW 300.0 i Evcass ua_vE eL s 100.0 N M 200.0 1 ~

w 5 _

W r

50.0 100.0

- 3 0 0 0  :

O' '

, 0.0 , , , , , , . , _ , [, ,, , , ,,

D. 3 5. 0 0. 0 5.O TIME (SECONOS3 T!*E (SECONOS) 1 LEVEL (INC4-REF-SEP-SKR T) 1 VOth AE) IVITY 2 VESSEL STEAMFLOW 3 TURBINE STEAMFLCW OPPLER ACTIVITY 200.0 t FEErwn.TF FLCu 1.0 3. SCRAM, REA TIVITY rgy _ 271 r,uyrv 100.0 _W.# A O. 0

~'

[

w N e

c0 ] ,

a

., a .

-1.0

-100.0

-2.0 C. 0 5.0 9.0 5.O TIME (SECCNDS)

TIME (SECONOS)

Figure 5. Plant Response to MSIV Closure i

12

22A8559 R:v. 0 AF ATURAL C: RCULATIO 1 B1 05 PERCEN T ROD LI 9E CL LTIMATE STABILITY LINE 1.00 C C A

x .75 N

(N X

C$

~

F-i

.50

>- t

=<

L)

LLI O

.25 --

r 0.00

0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 6. Reactor Core Decay Ratio 13 l

.r 22A8559 Rev. O x i

0. 0

-5.0

-10.0 / ,Ad

-15.0 V

H {l

(

o -20.O ,

t .

it 8-25.0 /

m e /

8-30.O O

r

-35.O 'i ci, c, n ircn vii necni n BUhEUULAT56 VALU$ 5U5~

C BOUND- VAL 280 CAL /G COLD DBDUND VAL 280 C %L/G HSB

-40.0

0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG-C.-

Figure 7. Fuel' Doppler Coefficient in 1/A*C j 14

22A8559 Rev. 0 20.O A ACCIDENT FUNCTION 8 BOUNDING VALU E 280 CAL /G 17.5 15.O m

o 180C U 12.5 z

-< --A A A 10.O a

w

- 7.5 s

o

, s 5. 0 2.5 0.~ 0

0. 0 5. 0 10.0 15.0 20.0 ROD POSITION, FEET OUT Figure 8. Accident Reactivity Shape Function, Cold Startup 15 i

22A8559 R:v. 0 20.O A ACCIDENT FUNC TION 8 BOUNDING VALUE 280 CAL /G 17.5 15.0 , .. ..

pCC O W 12.5 M

F-gj 10.0

.O f I

U f

7.5 l U

E 5. 0 -

2. 5.

0. 0 L_

0. 0 5.0 10.0 -15.0 20.0 ROD POSITION, FEET OUT Figure 9. Accident Reactivity Shape Function, Hot Startup 16

22A8559 Rsv. 0 40.O A SCRAM FUNCTION B BOUNDI NG VALUE '280 CAL /G 35.0 g 30.0 I

w 25.O s

_.J w

a m 20.0 m e /

w Z

v g 15. O s

w s

o o 10.0 w

e 5.0

0. 0 ,s, m .a

,, m,. -

0. 0 1.0 2. 0 3. 0 4.0 5.0 6.0 ELAPSED TIME, SECONDS Figure 10. . Scram Reactivity _ Function, Cold Startup 17

22A8559 Rev. 0 50.O A SCRAM FUNCTION B BOUNDI NG VALUE 280 C AL/G 40.O S

I w

E

< 30.O

_.J l w

CD m

CD LU Z

20.0 m N

6--

~

s F--

U

$ 10.0

/

0. 0 . .

m 0.0 1.0 2. 0 3.0 4.0 5.0 6.0 ELAPSED TIME, SECONDS Figure 11. Scram Reactivity Function, Hot Startup 18

22A8559 Rsv. 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 plan-ning and is comprised of the fuel bundles designated in Item 2 of this sup-plemental submittal. It in turn is based on the beat 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 shut-down 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 4

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 multiplication of the core, uncontrolled, fully controlled, and with the strongest rod out) were calculated at.the Beginning-of-Cycle 6 core average exposure corresponding to the minimum expected End-of-Cycle 5 core average exposure. The core was assumed to be in a xenon-free condition.

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 at BOC and the maximum calculated strongest rod out eff k,gg at any exposure point.- The strongest rod out k,gg at any exposure point-is equal to or less than:

19

. i i .

_ . m . . . . . . . .

22A8559 Rev. O k SR0 = (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 subcritical 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,SRO gg given in Table A-1 from the critical k,gg of 1.0, resulting in a calculated cold shutdown margin of 1.5% 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 transient) to a subcritical 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, xenoa-free, all rods out conditions at the exposure point of maximum cold reactivity with the soluble boron concentration 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. The results of the SLCS evaluation are given in Table A-2.

20 1

22A8559 Rev. 0 Table A-1 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL ROD WORTHS - NO VOIDS, NO XENON, 20*C Uncontrolled, K 1.117 Fully Controlled, K 0.950

~

Strongest Control Rod Out, K 0.982 R, Maximum Increase in Cold Core Reactivity 0.003 With Exposure Into Cycle, Ak Table A-2 STANDBY LIQUID CONTROL SYSTEM CAPABILITY Shutdown Margin (Ak) ppm (20*C, Xenon Free) 600 0.023 4

s 21

22A8559 R v. O REFERENCES 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 LWRs," TVA-TR78-02A, 1978.

A-3. " Verification of TVA Steady-State BWR Physics Methods," TVA-TR79-01A, 1979.

'I 22-t

____ . _ _ _ _ m

1 22A8559 RIv. O APPENDIX B PLANT PARAMETER DIFFERENCES Only 12 of the 13 safety / relief valves were considered operable. The capacity was 78.1% at a reference pressure of 1123 psig.

i 23/24

22A8559- Riv. O APPENDIX C INCREASED CORE FLOW The licensing analyses for Cycle 6 were done with a core flow of 105%

of rated flow which will bound operation at rated conditions.

The conclusions regarding LOCA analysis, reactor internals pressure drop, and flow-induced vibration as discussed in Reference C-1 are applicable to Cycle 6.

The flow-biased instrumentation for the rod block monitor should be signal clipped for a setpoint of 106% since flow rates higher than rated would otherwise result in a ACPR higher than reported for the rod withdrawal error.

REFERENCE:

C-1. " Safety Review of Browns Ferry Nuclear Plant Unit No. 1 at Core Flow Conditions Above Rated Flow During Cycle 5", NED0-22135, May 1982.

n 25/26 (FINAL)

.. I