ML20236G659

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Supplemental Reload Licensing Submittal
ML20236G659
Person / Time
Site: River Bend Entergy icon.png
Issue date: 07/31/1987
From: Charnley J, Lambert P, Zarbis W
GENERAL ELECTRIC CO.
To:
Shared Package
ML20236G634 List:
References
23A5819, 23A5819-R, 23A5819-R00, NUDOCS 8708040339
Download: ML20236G659 (17)
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Text

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l 23A5819 Revision 0 Class 1 July 1987 i

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL '

FOR RIVER BEND STATION RELOAD 1 Prepared: b P. A. bert ,

Verified:

W. A. bis Approved

  • M

. . C M ey?' Manager F el Licensing NUCLEAR ENERGY BUSINESS OPERATIONS

  • GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GEN ER AL h) ELECTRIC f.. 1/2 8708040339 870731 -

PDR ADDCK 05000458 P PDR

_-_______L

23A5819 Rev. O IMPORIANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Gulf States Utilities Company (GSU) for GSU's use with the U.S. Nuclear Regulatory Commis-sion (USNRC) for amending GSU's operating license of the River Bend Station.

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 informa-tion in this document are contained in the contract between Gulf States Utilities Company and General Electric Company for nuclear fuel and related services for the nuclear system for River Bend 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 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 mcy 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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9 23A5819 Rev. 0

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g ACKh0kLEDGMENT The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Submittal were performed by P.K. ku of the Nuclear Fuel and Engineering Services Department.

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23A5819 Rev. 0

1. PLANT-UNIQUE ITEMS (1.0)*

l Appendix A: Analysis Conditions l

Appendix B: Basis for Analysis of Loss-of-Feedwater Heating Event f Appendix C: Application of GEMINI Methods

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

Fuel Type Cycle Loaded Nitmber Irradiated BP8 SRB 094 1 20 BP85RB163 1 120 BP8 SRB 248 1 280 BP8 SRB 278 1 40 New BP8 SRB 299 2 11 2 BP8 SRB 305 2 52 Total 624

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 8,685 mwd /MT Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 8,465 mwd /MT Assumed reload cycle core average exposure at end of cycle: 15,533 mwd /MT Core loading pattern: Figure 1

  • ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. A letter "S" preceding the number refers to the U.S. Supplement, NEDE-24011-P-A-8-US, May 1986.

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23A5819 Rsv. 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)

Beginning of Cycle, K-effective Uncontrolled 1.108 Fu'11y Controlled 0.936 Strongest Control Rod Out 0.973 l

R, Maximum Increase in Cold Core Reactivity with 0.015 l Exposure into Cycle, Delta K

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3).

Shutdown Margin (AK) jggg (20'C, Xenon Free) 660 0.04 6.. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (5.3.2.1.5 AND S.2,2)

(REDY Events Only) ,

Void Fraction (%)

  • Average Fuel Temperature (*F)
  • Void Coefficient (4/% Rg)
  • Doppler Coefficient (d/*F)

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  • See Appendix B.

8 * *~

YMinh _ ~ _ _ _ .

23A5819 Rev. 0

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)

Exposure: . BOC2 to E002 Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) _

(1000-1b/hr) MCPR BP8x8R 1.20 1.53 1.40 1.051 6.941 107.9 1.14

8. SELECIED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: Yes Rod kithdrawal-Limiter: Yes Thermal Power Monitor: Yes Improved Scram Time: No Exposure Dependent Limits: No Exposure Points Analyzed: EOC l

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single-Loop Operation: Yes I- Load Line Limit: No Extended Load Line Limit: No Increased Core Flow: No Feedwater Temperature Reduction: ho Feedwater Heaters Out-of-Service: Yes I ARTS Program: No Maximum Extended Operating Domain: No l

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23A5819 Rev. 0

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

Flux Q/A ACPR Transient (% NBR) (% NBR) BP8x8R Figure Exposure Range: BOC2 to EOC2 Load Rejection w/o Bypass 286 108 0.07 2 Lor,s of 100*F Feedwater *

  • 0.11
11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

(

SUMMARY

S.2.2.1)

The generic bounding BWR/6 Rod Withdrawal Error analysis described in NEDE-24011-P-A-8-US is applied; the resulting aCPR is 0.11.

12. CYCLE MCPR VALUES (S.2.2 and S.2.5.4)**

Non-Pressurization Events Exposure Range: BOC2 to E002 BP8x8R Loss of 100*F Feedwater Heating 1.18 Rod Withdrawal Error 1.18 Pressurization Events Exposure Range: BOC2 to E002 BP8x8R Load Rejection Without Bypass 1.15 Feedwater Controller Failure 1.14

  • See Appendix B.
    • GEMINI ODYN adjustment factors are provided in the letter from J.S. Charnley (GE) to M. W. Hodges (NRC), " GEMINI ODYN Adjustment Factors for BWR/6", dated July 6, 1987.

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23A5819- Rev. 0

13. 'OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) s1 y Transient (psig) (psig) Plant Response MSIV Closure 1210 1247 Figure 4 (Flux Scram)

14. CONTROL P.0D DROP ANALYSIS RESULTS (S.2.5.1)

Banked Position Withdrawal Sequence is utilized at the River Bend Station'; therefore, the bounding Control Rod Drop Analysis (CRDA) described in NEDE-24011-P-A-8-US is applied. NRC approval of the bounding analysis is given in the letter to J.S. Charnley (GE),

" Acceptance for Referencing of Licensing Topical Report Amendment 9 to NEDE-24011, Revision 6, "GESTAR-II General Electric Standard Application for Reactor Fuel," January 25, 1985.

15. STABILITY ANALYSIS RESULTS (S.2.4)

GE SIL-380 recommendations have been included in the River Bend Station aperatin8 Procadures and/or Technical Specifications; therefore, no stability analysis is required.

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

LOCA Method Used: SAFE /REFLOOD (see River Bend Station Final Safety i

Analysis Report)

Fuel Type: BP8 SRB 299 Average Planar Exposure MAPLHGR 0xidation (GWd/ST) (GWd/MT) (kW/ft) PCT (*F) Fraction  ;

l 0.20 0.22 11.5 2019 0.016 l 1.0 1.1 11.6 2020 0.016 5.0 5.5 11.9 2019 0.016 10.0 11.0 12.1 2035 0.016 15.0 16.5 12.3 2067 0.018 20.0 22.0 12.3 2081 0.019 25.0 27.6 11.8 2035 0.016  ;

35.0 38.6 11.0 1905 0.010 l 45.0 49.6 9.6 1725 0.005 50.0 55.1 8.7 1625 0.003 11 L -

23A5819' Rev. 0 Fuel Type: BP8 SRB 305 .)

Average Planar. Exposure MAPLHCR 0xidatiori. i

-(GWd/ST) (GWd/MT)' (kW/ft) PCT (*F) Fraction. -l 0.20- 0.22 11.4 2018 .0.016 '

=1.0 1.1 11.5 2014 0.016 5.0 ~5.5 11.8 2006 0.015 10.0 11.0 12.6. 2092 0.019

'15.0 16.5 12.7 2121 0.021 l 20.0 22.0- 12.7' 2134, 0.022 l 25.0: 27.6 12.1 2064 0 018 35.0 28.6 10.9 .1902 0.010 45.0 49.6- 9.4 1707.- 0.005 50.0 55.1 8.5 1609 0.003~

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23A5819 Rev. 0 1

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l l l ll III 1 357 91113151719212325272931333537394143454749515355 FUEL TYPES A = BP8 SRB 094 D = BP8 SRB 163 3 B = BP8 SRB 248 E = BP8 SRB 305 {

! C = BP8 SRB 278 F = BP8 SRB 299 Figure 1. Reference Core Loading Pattern l ~

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23A5819 Ray. O 1 NEUTRON FLU ( l VESSEL PRESS RISECPSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET 7 LOW 3 RELIEF VALVE FLOW

. 13 8, 0 300.0 e ove* SS u ni ug r! nu _

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1 LEVEL (INCH-REF-SEP-SKRT3 1 VOID REACTIVITY 2 VESSEL STEA1 FLOW 2 DOPPLER REACTIVITY -

3 TURBINE STEAMFLOW 3, SCRAM REA,CTI, VI, T Y 200.0 _. i e re_ n_ u i t. r o..e_ n. u_

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0. 0 2.0 4.0 6.0 0.0 2. 0 4.0 TIMF (SECONDS) TIME (SECO C S)

Figure 2. Plant Response to Generator Load Rejection Without Bypass 14

23A5819 Rev. 0 150.0 1 NEUTRON FLU 1 VESSEL PR SS RISE (PS])

2 AVE SURS ACE AT FLUX 2 SAFETY V VE FLOW 3 CORE INLET OW 3 RELIEF V VE FLOW 150.0 'reoE 'M E' '9 4 BYPASS V/L'E FLOW

.k 100.0 y 4 d-100.0  : r NT _

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t +4 0.0 _

.. . . . _ . ., 1,. g m'_y 0**

0. 0 10.0 20.0 30.0 0. 0 10.0 20.0 3 TIME (SECONDS) TIME (SECONDS) 1 LEVEL (INCH-REF-SEP-SKRT) 1 VOID REACdIVITY 2 VESSEL STEA1 FLOW 2 DOPPLER EA:TIVITY .

3 TURBINE STEAMFLOW c V 150.0 ' errnuavro e nu 1.0 3, SCR env AM er r r,I, u ,I ,T u R y Y l

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100.0 22 2 2* + ~ Lp k 6.0 ,~

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Figure 3. Plant Response to Feedwater Controller Failure I

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23A5819 Rev. 0 1 NEUTRON F.UX 2 AVE SURFA:E HEAT FLUX 1 VESSEL PRESS RISE (PSI) 3 CORE INLET FLOW 2 SAFETY VA.VE FLOW 150.0 3 RELIEF VALVE FLOW 300.0 e. evo *SS uniuE etow S.

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200.0 5

kk I

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0. 0 . , 0. 0 _ , _ , . . ., ,, .
0. 0 5.0 0. 0 5.0 TIME (SECONDS)

TIME (SECONDS) 1 LEVELCINC4-REF-SEP-SKR T) 1 VOID EACTIVITY 2 YESSEL STEAMFLOW 3 TURBINE STEAMFLOW 2 00PPL REACT!v!TY -

2 n n,0 e rggruivre rLgu 1.0 SCRA EACTIVITY- i 3, vgy,Mt gigt , v , w n o,0 .m--. .

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TIME (SECONDS)

Figure 4. Plant Response to MSIV Closure (Flux Scram) 16

23A5819 Rev. O APPENDIX A ANALYSIS CONDITIONS To accurately reflect actual plant parameters, the values shown below were used instead of the values in NEDE-24011-P-A-8-US.

Non-Fuel Power Fraction 0.038 Pressure Relief System:

No. of Valves 16

/

Lowest Safety Setpoint (psig) 1165 Lowest Relief Setpoint (psig) 1103 Capacity (ib/hr) 869,000 at 1130+3% (psig) kg 17/18

23A5819 Rev. O APPENDIX B BASIS FOR' ANALYSIS OF LOSS OF FEEDWATER' HEATER EVENT The Loss of Feedwater Heater event was anslyzed with the 3D BWR Simulator code. described-in NEDE-24011-P-A-8-U3. The transient analysis inputs normally reported,in Section 6 of the licensing submittal are internally calculated in the 3D BWR Simulator code and in ODYN. The transient plots, flux, and Q/A normally reported in Section 10 are not outputs of the 3D BWR Simulator code.

'efore, these items are not included in this document.

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23A3819 R:v. O APPENDIX C APPLICATION OF GEMINI METHODS The GEMINI system of methods are used to perform the licensing analyses j of River Bend Station (RBS) Reload 1. The GEMINI system of methods is described in Reference 1; NRC approval of these methods is documented in Reference 2. In Reference 3, the application of GEMINI methods in licensing analyses is described. Pressur1zation events that could establish the Operat-ing Ilmit MCPR (Load Rejection Without Bypass and Feedwater Controller Failure) cre analyzed at the 100% power level. Power level uncertainties specified in Regulatory Guide 1.49 are accounted for by adding adjustment factors to the calculated ACPR. NRC approval of this procedure is provided in Reference 4.

Rod Withdrawal Error The NRC approved generic Rod Withdrawal Error analysis for BWR/6's d: scribed in Reference 5 is applied to RBS Reload 1. An evaluation of the impact of GEMINI methods on the generic analysis indicates that the results of the generic analysis continue to be conservative and bounding.

Overpressurization Analysis The MSIV Closure (Flux Scram) analysis is performed using GNMINI methods at the 102% power level to account for the power level uncertainties specified in Regulatory Guide 1.49.

l Control Rod Drop Accident t

f The NRC approved bounding Control Rod Drop Accident analysis for Banked Position Withdrawal Sequence plants (such as RBS) described in Reference 1 is applied to RBS Reload 1. The impact of GEMINI methods on the results of the generic analysis is negligible.

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23A5819 Rsv. O Stability The NRC approved generic stability approach described in Reference 1 is applied to RBS Reload 1. The use of GEMINI methods does not impact the generic analysis.

References

1. Letter, J. S. Charnley (GE) to C. O. Thomas (NRC), " Amendment 11 to GE LTR NEDE-24011-P-A," February 27, 1985.
2. Letter, C. O. Thomas (NRC) to J. S. Charnley (GE), " Acceptance for Refer- ,

encing of Licensing Topical Report NEDE-24011-P-A, Rev. 6, Amendment 11,

' General Electric Standard Application for Reactor Fuel'", November 5, 1985.

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3. Letter, J. S. Charnley (GE) to H. N. Berkow (NRC), " Revised Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE-24011-P-A", January 16, 1986.
4. Letter, G.'C. Lainas (NRC) to J. S. Charnley (GE), " Acceptance for Refer-encing of Licensing Topical Report NEDE-24011-P-A, 'GE Generic Licensing Reload Report', Supplement to Amendment 11", March 22, 1986.
5. "GESSAR-II 238 BWR/6 Nuclear Island Design", NRC Docket No. STN50-447.

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