ML18025B054

Application for Amend of License DPR-68,changing Tech Specs to Accomodate Reload 3,Cycle 4 Operation of Unit 3.Shutdown Planned for 801020 to Begin Refueling Outage W/Restart on 801201.Class III Fee Encl

ML18025B054
Person / Time
Site:	Browns Ferry
Issue date:	08/27/1980
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18025B055	List: ML18025B054 ML19344E653 ML19344E654
References
TVA-BFNP-TS-148, NUDOCS 8009020406
Download: ML18025B054 (30)

Text

Y1003J01A03 Class I August 1980 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY NUCLEAR POWER STATION UNIT 3 RELOAD NO. 3 Prepared:

C. L. Hilf Approved:

R. E. Engel, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION ~ GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA95125 GENERAL ELECTRIC DOQOZ, c) QC3$

Y1003JOlA03 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT 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 Unit 3. 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 The Tennessee Valley 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,,

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 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.

Y1003JOlA03

1. PLANT-UNI UE .ITEMS 1.0

• Items different from or not included in Reference 1:

7'ata 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 F~uel T e Number Number Drilled Initial Core 8DB219 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 l.

Y1003 J01A03

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CORE SYSTEM 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 LI UID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (Ak)

(20'C, Xenon Free) 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 Average Fuel Temperature ('F) 1343 Scram Worth N/A ($ ) -37.67/-30.13 Scram Reactivity vs Time Figure 2

7. RELOAD-UNI UE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5 ')

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 '.526 Bundle Flow (10 lb/hr) 108.2 108.5 109.2 Initial MCPR 1.24 1.25 1.25

• N = Nuclear Input Data A Used in Transient Analysis

Y1003J01'A03

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5. 2. 2)

Recirculation. Pump Trip

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.2)

Core P

'Power Flow 4 Q/A el V 4CPR Plant Transient e p e (2) (2 NBR) ( . NBR) (psig) (psig) Sx8 8xSR PgxBR Response e

Load Refection BOC4-EOC4 104.5 100 239 111 1226 1250 0.17 0.18 0.18 Figure 3 Without Bypass e Loss oE 100 F 104.5 100 124 123 1013 1069 0.15 0.15 0.15. Pigure 4 Feedwater Heating Feedwster BOC4-EOC4 .,104 ' 100 164 112 "

1155 1189 0.12 0.12 0.12 Figure 5 Controller Failure

10. LOCAL ROD .WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(5.2.1) b CPR~~ XLRGR Rod Block Rod Position 8x8R/ 8x8R/ Limiting

~Readia (Feet Withdrawn) 8x8 P8x8R 8x8 '8xRR Rod Pattern 104 3.5 0.10 14.8 Figure 6 105 4.0 0.12 0.11 15.2 16.5 16.2'5.2

'igure 6 106* 4.5 0.14 , 0.12 16.6 Figure 6 r

107 108 7.5'.10 4.5 5.5 0.14 0.18 0.12 0.14 ,

15,2 15.2 16.6 16.7'5.2 Figure Figure 6

6 109 6.5 0.20 0.16 16.7 Figure 6 110 0.20 0.17 15.2 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 l.

+*+A 2.2X 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.

Y1003J01A03

11. OPERATING MCPR LIMIT (5.2),

BOC4 to EOC4 1.24 8x8 Fuel

1. 25 8x8R Fuel

l. 25 P8x8R Fuel

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5. 3)

Power Core Flow 'sl P V Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 104.5 100 1265 1299 Figure 7 (Flux Scram)

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability:

Decay Ratio, *x /x 0.85 (105% Rod Line - Natural Circulation Power)

Channel Hydrodynamic Performance Decay Ratio, x /x (105% Rod Line Natural Circulation Power) 8x8R/P8x8R 0.29 8x8 0.36

14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)

Reference 2.

Y1003J01A03

15. LOADING ERROR RESULTS (5.5.4)

Limiting Event: Rotated Bundle PSDRB265L MCPR: 1.08"

16. CONTROL'ROD'DROP ANALYSIS RESULTS (5.5.1)

Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and ll Scram Reactivity Functions: Figures 12 and 13 Plant specific analysis results Parameter not bounded: None

Y1003J01A03 REFERENCES

1. General Electric Boiling Water Generic Reload Fuel Application, NEDE-24011-P-A, August 1979.

2. Loss-Of-Coolant Accident Analysis Report for Browns Ferry Nuclear Plant Unit 3, NED0-24194A, July 1979.

3. Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 3 Reload 1, NEDO-24128 (Appendix A), June 1978.

Y1003J01A03 60 PFQA P8 QA QAI I Ipe PAIQF 58 g K EI Do EJ Do EJ EJ lm Oo OB 56 Qe De pcIQO DH OH ph Qo DH,QHQQ QH QH po,pc OA,OG 08 54 ps OD pcpopE pe Qe Oo OEQG CGlOE Qo OG QG pE Qo Qc Qo OB

~

52 5o 48 08 Qe K El K El OA OA K K po K pc po OG Qo DAI Os IJ K El K

I Oo QG DA DD QEIK IDA Do po COIIQE DCIQA OoIOA OE po Kl RIIJ K El DGIDG KIDD DGIQG Qo CCI OE Qo DIOG K K 08 Im QA 46 KK CCI Qo OE Pe Os K pePo Qe PE CCI ps DG PE Os Pe Do DE Oe Qs 0+CI Pc+Oh K 44 Ph Dc EJ Q EI KI Ph EIKIEJIK EJIK DA [P EJ I Do [Q PAI 42 Po Col QE Oo Pc OE Pc K DA K +Pc Pc Do+DE K+DO Z 40 Q8 Qo OH DG KJ EJ OE DG KI De Q Dc Oo Ds EJ g] [g Dc K CCI KI Po Im El 38 PA'Pc 8 OG Csl Oo'OE ps De Pc DG DG K

QG QG Qo K

QG DH 8

36 DAIIQDDAIIKK KIDAEJEKIEIKKJEIDCEODOAEJtmZ Kl QE Pe Cel OE Ps gs PE Oe Cel CCI Os Os Col Qe Pc 34 Ipc Qo QE ~ pc Qo pc Qc po pc K pc 32 Z Po OH Oe OE Qo QE CGJ Ps DE Pe 06 OE PszOG OGIOG IK Qc QE Cel Qo 8 K 30 Oo QH Os KIIJ ps Os Qe Qe ps C<IEJ KI OE ElKI QGIOG OOIQE Do DE QE Qs Qs Oe Qs Qo QE Oe OH po ph 28 26 24 ph pc QOIQE Z [I El Kl El E K El EI pC pH Qe Qo OE QE OH pG Dc pC OE Qo lm Qs pc K

I I

QGjpe DCI IQo IQc IQE I Kl K El E El Z E El K E EI IXI EI Kl lm Al IOE OE~

IAI QEIQO pC KIDD pc KIQO QejQH I

[g IJ g) E EI Im QG QG QG QG QG 22 QH DE pe pc Qo IJ fg [g Kl Kl IJ Q IJ Im 20 Qo+K + PE+ Pc+ Pc+K QG QG Pc+ +Pc Qo DG QG Qo

+

OH Qo 18 QA+Qo Dc+Pc DE+Pc +Pc Do+DE PE+Do Po+

PAPAEIEIKEIKIKKKKIKIEIDAKJKlEIOD CCIKZ 16 pA QA Qo OE pe Qs K Qo Qe Qs K Qs Qe QE Qe, Qe Qo, QE pe QEQO pcIKA 14 Z EJ KIQo DG ps QD CEI OG pe po OEQG OGK DD QG OG KQD DG Csl Qo 12 08 Z KK po OE Dc Col OE Col Qo pc K Oo pc DE Oo QE K 08 10 pa+pA pA+pa p+pc pa+pa pA+pc pa+pa pc+p~ pa+pa pc+p~ pa+pA E+pa 8 EIEIEIEIKOe Oe Col KIOG lE K EIE Oe EIEI El Q8 6 0808CAIDCQO880AOOQHOHOO880oCCIK0808 4 ggKIEJ EJ IJEJOOODEJ IIEJ II,EIKIBJ 2 p+pA pa+CA] p+Z pA+Z pA+pA pA+pa pA+pF I I I I I I I I I I I I I I

'1 3 5 7 S11 131517 1921 2325272S313335373941 4345474S5153555759 FUEL TYPE.'

AR SDB219,IC' ~ PSDRB265L,R3 SDB219,IC F w SDB219 IC C w SDRB265L,R1 G ~ BDB219,IC D ~ PSDRB265L,R2 H ~ SDRB265L,R1 a IC ~ INITIALCORE, RI ~ RELOAD 1, R2 ~ RELOAD 2, etc.

Figure 1. Reference Core Loading Pattern

Y1003J01A03

.100 C-679 CRD IN PERCENT I-NOHINRL SCRRH CURVE IN (-0) 90 2-SCRRN CURVE USEO IN RNRLl'SIS 40 80 35

'0 30 60 Q

25 50 (A

o 20 40 15 30 10 20 10 0 0 0 2; 3 TIME (SECQNDS)

Figure 2. Scram Reactivity and Control Rod Drive Specifications

1 NEUTRON LUX 1 VESSEL ES RISE IPSI) 2 AVE SURF CE HEAT FLUX 2 SAFETT V LVE FLOW 150. 3 CORE INL T F OH 3 RE EF V LVE FLOW 8 BTPRSS V LV LOH 5

6 o 100.

I I

100.

0. 0.

0. 8. 12. 16. 0. 8. 12. 16; TINE (SEC) TINE (SEC)

I LEVEL(1 H-REF-SEP-SKIRT I VOIO AE TIVITT 2 VESSEL 5 EAHFLOH 3 ~TUR8INE TEANFLOH 3 SCRAN RE CTIVITT 100.

0.

-100.

0 8. 12. 16. 0. O.Q 0.8 1.2 1.6.

TINE (SEC) TINE ISEC)

Figure 3. Plant Response to Generator Load Rejection Mithout Bypass

I VESSEL P ES RISE (PSI)

"I HEUTROH VC~hr LUX GE-HEAT-R.UX 2 RELIEF V LVE FLOH 3 CORE IMl I FLOH 125.

3 BTPASS Vl V F 150. 'l CURE IHL I SUB 5

o 4J 100.

I h

I Ps 50.

0. -25.

0. QO. 80. 120. 160. 0 QO. 80. 120. 160.

TIME (SEC) TIME (SEC)

I LEVEL(I H-REF"SEP-SKIRT I VOID REA T I VITY 2 VESSEL S EAMFLOH 2 OOPPLER EACT I V ITT 3 TURBINE TEAHFLOH 3 SCRAH AE CTIVITT 150. CTIVITV th Vl 100. I 0.

W 0

LJ

-1.

I 0.

0. 80. 120. 160. 0 )(0. 80. 120. 160.

TIME (SEC) TINE (SEC)

Figure 4. Plant -Response to Loss of 100 Deg F Feedwater Heating

I NEUTRON LUX 1 VESSEL P ES RISE IPSI) 2 AVE SURF CE HEAT fLUX 2 SAFETT V LVE FL%

150. 3 CO I FLOW 125. 3REI FV V FOH 9 CORE INL I SUB 0 BTPRSS V LV LON 5 5 6

oUJ 100.

I I

Pu 50.

g 0.

0. 10. 20 0 30. 40. 10. 20. 30. QO.

TINE (SEC) TINE (SEC)

I LEVEL (I -REF-SEP-SKIRT 1 VOIO RE T IVIT 2 VESSEL S EAHFLOK 2 OPPLER ERG ITT 3 TURBINE TEAHFLOW CRAH RE CT ITT 150.

100.

0.

0. 10. 20. 30. 40. 0 12. 18.

TINE ISEC) TIRE (SEC)

Figure 5. Plant Response to Feedwater Controller Failure, Maximum Demand

Y1003J01A03 1

NOTES: lo ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC (FULL CORE SHOWN)

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 55 36 36 36 36 ., 36 51 6 6 2 2 ~ 6 6 47 36 36 36 36 36 36 36 43 39 35 31 27 6

4 4

36 36 2,

6 2

36 36 10 14 14 36 40 14 0

0 40 36 14 0

36 40 10 14 14 36 36 6

2 2

36 36 6

4 4

23 36 36 36 40 36 36 36 19 6 6 10 14 10 6 6 15 36 36 36 36 36 36 36 v

ll 7

6 36 6

36 2

36 2.

36 36 6

3 4 Figure 6. Limiting RWE Rod Pattern 12

1 NEUTRON LUX 1 VESSEL P ES RISE (PSI) 2 AVE SURF CE HEAT FLUX 2 SAFETT V LVE FLON 150. 3 CORE INL T F OW 3 RELIEF V LVE FLON 0 BTPASS V LV ON 5

6 r) 100.

I Pu 50. 100.

0. 0.

0 8. 12. 16. 0. 8. 12. 16.

TIHE (SEC) TIHE (SEC)

I LEVEL(I H-REF-SEP-SKIRT 1 VOIO RER TIVIT 2 VESSEL S ERHFLON 2 OGPPLER V ITT 3 TURBINE TEAHFLON 3 SCR CTIVITY 100.

0.

-100.

0 8. 12. 16. 0. 0.6 1.2 1.8 2. (I TINE (SEC) TIHE (SEC)

Figure 7. Plant Response to MSIV Closure

Y1003J01A03 1.2 ULTIMATESTABILITYLIMIT 1.0 O

0,8 X NATURAL O CIRCULATION I-K 0.6 105% ROD LINE 0,4 0.2 0

0 40 60 80 PERCENT POWER Figure 8. Decay Ratio 14

Y1003J01A03 A CALCULATEDVALUE@OLD B CALCULATEDVALUEWSB C BOUNDING VALUE FOR 280 eel/g, COLD D BOUNDING VALUE FOR 280 eel/g, HSB

-10 xI-z 0

-16 e

z III O

IL -20 IL UI 00 L

III 0 -26 0

-30 0 1000 1600 2000

, FUEL TEMPERATURE tdeg C)

Figure 9. Doppler Reactivity Coefficient Comparison for RDA

Y1003J01A03 A ACCIDENT FUNCTION B BOUNDING VALUE FOR 280 eel/II 15 10 0

0 10 15 20 ROD POSITION, feet OUT Figure 10. Accident Reactivity Shape Function at 20 C

Y1003JOlA03 20 A ACCIDENT FUNCTION B BOUNDING VALUE FOR &0 csl/g, Z

I 0

Z 0

X I-Y 10 0

b K

0 0 10 15 ROD POSITION, fest OUT Figure 11.. Accident Reactivity Shape Function at 286'C 17

Y1003J01A03 60 A SCRAM FUNCTION B BOUNDING VALUE FOR 280 ceI/O 40 x

O D

0 x

I 30 5C 0

E3 z

20 I-I-

O K

10 0

0 ELAPSED TIME, seconds Figurerl2. Scram Reactivity Function at 20'C

Y1003J01A03 A SCRAM FUNCTION 75 B BOUNDING VALUE FOR 280 eel/g 50 25 0

0 3 4 ELAPSED TIME, seconds Figure 13. Scram Reactivity Function at 286'C 19/20

I C

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'urn is based on the best possible prediction of the core condition at the end of the present cycle and on the desired c'ore energy capability for the reload cycle. It is designed with the intent k

that it will represent, as clos'ely 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-l, 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.l. 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.

Y1003J01A03 Cold keff was calculated ff with the strongest control rod out at various exposures through the cycle. The value R is the difference between the strongest rod out keffff at BOC and the maximum calculated strongest rod out ff at any exposure point. The strongest rod out keff keff ff at any exposure point is equal to or less than:

k SRO = (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 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 SRO by subtracting the keff given in Table A.l from the critical keff of 1.0, resulting in a calculated cold shutdown margin of 1.1 percent Ak.

A-2

Y1003JOlA03 Table A.'1 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL ROD WORTHS NO VOIDS, NO XENON, 20 C UNC Uncontrolled, K ff eff 1.120 CON Fully Controlled, K eff 0.955-SRO Strongest Control Out, Rod K ff eff 0.989 R, Maximum Increase in Cold Core Reactivity 0.000 With Exposure Into Cycle, Ak A-3

Y1003J01A03 REFERENCES A-l. 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 LWR's," TVA-TR78-02A, 1978 A-3. "Verification of TVA Steady-State BWR Physics Methods," TVA-TR79-01A, 1979 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 Rated Steam Flow: 14.09 x 10 6 lb/hr GETAB Analysis Initial Conditions Reactor Pressure: 1035 psia Inlet Enthalpy: 521.5 Btu/lb king penalty for densification is included.

one safety/relief vlave out of service.

B-1/B-2

N I

'E I

C 4 ~ I r l

Y1003JOlA03 APPENDIX C MARGIN TO SPRING SAFETY VALVES The rationale for changing the basis for providing pressure margin to the safety valves is presented in Reference C-1. This change has been h

'pring accepted by the NRC (Reference C-2).

On this basis the plant can operate at full power throughout the cycle.

Y1003 J01A03 REFERENCES C-l. J. F. Quirk (GE) letter to Olan D. Parr (NRC), "General Electric Licensing Topical Report NEDE-24011-P-A, 'Generic Reload Fuel Submittal," dated February 28, 1979.

Application,'ppendix D, Second C-2. Letter, T. A. Ippolito (NRC) to D. L. Peoples (Commonwealth Edison Co.)

enclosing a Safety Evaluation supporting Amendment No. 42 to Facility Operating License No. DPR-25.Dresden Nuclear Power Station Unit 3, dated April 16, 1980.

C-2